ML20059L496

BWR Owners Group NUREG-0578 Implementation: Analyses & Positions for Plant-Unique Submittals

ML20059L496
Person / Time
Site:	Browns Ferry
Issue date:	08/31/1990
From:	BWR OWNERS GROUP
To:
References
RTR-NUREG-0578, RTR-NUREG-578 NEDO-24782, NUDOCS 9009270061
Download: ML20059L496 (167)

Text

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BWR OWNERS' GROUP  !

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NUREG-0578 IMPLEMENTATION:  !

ANALYSES AND POSITIONS FOR  !

PLANT-UNIQUE SUBMITTALS i l

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, , NED0-24782 e 's .. TABLE OF CONTENTS Page -

1. INTRODUCTION 1-1

2. REVIrl 0F CONTAINMENT PER NUREG-0578 EQUIREMENT 2.1.4 2-1

3. DETINITION OF ESSENTIAL AND NON-ESSENTIAL SYSTEMS PER NUREG-0578 REQUIREMENT 2.1.4 3-1 1

4

4. SYSTEMS CONTAINING RADI0 ACTIVE FLUIDS

, PER NUREG-0578 REQUIREMENT 2.1.6a 4-1 l -. V EQUIPMENT REQUIRING ACCESS FOLLOWING AN ACCIDENT

! PER NUREG-0578 REQUIREMENT 2.1.6B 5-1 1 .

j 6. RADIATION SOURCE TERM INFORMATION .

'PER NUREG-0578 REQUIREMENT 2.1.6B 6-1 i

I 7.

BERCENERIC'SAMPLINGSYSTEMCONCEPTUALDESIGN 1 PER NUREG-0$78 REQUIREMENT 2.1.8A 7-1

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B. SUPPLEMENTARY INFORMATION ON REACTOR COOLANT SYSTEM VENTING PER NUREC-0578 IMPLEMENTATION LETTER REQUIREMENT ,

8-1 l

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9. INFORMATION REQUIRED IN TECHNICAL SUPPORT CENTER 9-1 APPENDICES A. CONCERNS RELATED TO PRIMARY CONTAINMENT ISOLATION LOGIC A-1

3. RADIATION SOURCE TERM INFORMATION PER NUREC;.0578 REQUIREMENT 2.1.6B J-l C. Eh'R LMNERS' CROUP POSITION ON H1Cll-POINT VENTS ' C-1

0. IIPf:1 TURl!!NE PERFORMANCE k'!Til IlYDROCEN IN SUPPLY STEAM D-t 4

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. ,,J ILLUSTRATIONS

,j Figure Title p,g, ,,

t i 2-1 MSIV Logic Change  !

2-8

. 2 Reiommended Design Change for so'lenoid/ Pilot Valves 2-9 2-3 Temporary Design Change for Solenoid / Pilot Valves 2-10 -

7-1 Pressurised RPV Liquid Sampler

• 7-2$ l 7-2 Low-Pressure or Depreesurized RPV Liquid Sampler 7-26  !

7-3 Suppression Pool Sample Line 7-27  !

7-4 Sample Room Layout Concept  !

7-28 7-5 Proposed Low-Pressure and Depressurized Sample Line Tie-In 7-29 j i-6 Dissolved Cas and Large-Volume Liquid Sampler 7-30 1 1-I f Small-Volume Liquid Sampler e. 7-3}'

784 Proposed Typical Gas Sample Line 7-32 -

7' Atmospheric Sampler '

7-33 7-10. Direct Dose Rates from Fission Products in (Pilgrim) i Reactor Building at 1 Day Post-LOCA 7-34

' 11 Plant Model of (Pilgrim) Radiation Transport Pathways 7-35

[ 7-12 Whole Body Dose Rate Versus Ground Level Distance frc

• Stack (Pilgrim) fSr Power = 1 MWt at 1 Hour Post-LOCA 7-36 7-13 Inhalation Dose Rates (Pilgrim) for Power = 1 MWt at  !

1 Hour Post-LOCA 7 37 i

,, TABLES

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Title Page 3- 1 Essential / Nonessential Equipment 3-2 '

4 . Systems Containing Highly Radioactive Pluids 4-2

, 7-1 Assumptions for Radiological Analyses 7-24 I W

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1. INTRODUCTION -

1 s  ;

' this report contains the results of studies conducted by General Electric Con- -

i 1

,  ; ,any to assist its utility customers in complying with certain provisions of i

.TRCC-0578 (the U. S. Nuclear Regulatory Commission's "short-tern Lessons -

. earned Report"), and the September 13, 1979, letter from D.G. Eisenhut (NRC) m all operating nuclear plants requiring the implementation of NUREG-0578.

"w work herein was authorized by the BWR Owners' Group in its November 9,

79, general meeting, and was funded by the owners' Group.

N'ne of the sections of this report are intended to serve as free-standing

('

• • k'onsestoanyoftheNUREG-0578 requirements;rather,theyarespecialized

..'ntributions which are to be used by individual utilities in preparing their en submittals. ,

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2. REVIEW OF CONTAINMENT

.h; PER NUREG-0578 REQUIREMENT 2.1.4 i

This section provides a generic approach to Requirement 2.1.4, " Containment

. 1;olation Provisions for PWR's and BWR's," of NUREG-0578. It is based on a j jetsiled review of one BWR owner's approach, and on dicsussions with other BWR i

..ners. Some identified problem areas which may be generic are identified, and 7 dternative solutions are proposed. Also identified are some concerns which I l a 4ve been raised which do not appear to require action. These, too, are of gen- l' rie interest. There may be plant-unique isolati'm concerns, of course, which

'.s v -

s iot been identified in this generic study. It is the responsibility of >

aca EWR utility to make it's own final findings and cosaritments for action.

'tEVIEW OF PENETRATIONS / ISOLATION PROVISIONS I

f

[ A review should be made of each containment penetration to determine the pres-I at isolation requ'irements and the, provisions for obtaining that , 01stion. The

. allowing types of documenta.may be used to assist in forming the .aasis for the

• r view: Schedule of Penetrations, Penetration Leak Rate Test Procedure, and

'mtsinment Isolation System Elementary Diagram.

i 1 e

i .' . 2 DETEPJtINE ISOLATION VALVE CLASSIPICATIONS E

The isolstion valves should be' classified according to valve types, as follows: ,

i A. Motor-Operated Valves v

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[ b. Solenoid-Operated Valves

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c. Air-Operated Valves with Pilot Solenoids l

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d. Testable Check Valves ,

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e. . Manually Operated Valves ji

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Nachanical (non-testable) Check Valves  !

2.3 DETERMINE TYPE OF OPERATING SWITCH USED i

The types of electrical switches 'tsad to control the remotely actuated isola-

• tion valves generally fall into one of the following categories. The appro- f, priate category should be determined for each isolation valve. I n >

a. Three position, "close-normal-open", sptina 'teturn '.o " normal". i I  !

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, b. O

v t Twoposition,"Close-open"or"close-auto /open",maintain

c. .I Two position, "close-op,en", spring return to "close". ,)

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d. Dual' pushbutton, "close" and "open". W

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.s 2.4 DETERMINE TYPE OF SYSTEM h"

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Generally, the line's penetrating the primary containment fall into two categor- ie) '

i ies, i.e. , open-loop or closed-loop. The open-loop designs are those .tystems '. E y ,

which are open to either the containment or the pressure vessel. -Ex . ,

- es.of ,-}

these are containment ventilation lines, reactor coolant sample'11e d RHR t

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process lines. Closed-loop designs are those systems which do not icate .

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l directly with the containment or reactor vessel, such as the neutt :or

.. 1 TIP lines and the recirculation rump bearing cooling water lines. t 2.5

• DETERMINATION OF VA1.VES REQUIRING MODIFICATION Generally, the types of valves requiring modification f all into o:

following two classifications: of ti}e i  !

da Solenoid Operated;

b. Air operated with Pilot Solenoid.

2-2

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.yfitionally, they can be desctibed as having two posittor, operating switches y sich maintained contacts and will be part of an open-loop system. "

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..o - EY.AMPLES OF VALVES REQUIRING M00.FICiTION 1

he following types of valves were found en one plant reviewed to require mod-' )

.i! cation. The numbers in parentheses are the number of such valves in the-

..iant reviewed. i e

l i de' Drywall Compressor (2J~

p h '-

h. Air ~ Purge Supply inlet (1) ..

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tet. ..

[ c. Dryvell Air Purge (1) f o

'd. Torus Air Purge (1) i.

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e.- Drywell ent (1),, ,

E h- f. Torus Vent (1)

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g. Drywell Vent to Reac' tor Building Exhaust System (1) s .

h. . ' Torus. Vent to Reactor Building Exhaust System (1)' '

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1. Drywell 2". Vent Relief (1)

J J. Torus 2?' Vent Relief (1), ,

k. Consainment Ventilation (2) 1.

Suppression Pool Vent Header (4) m ., Core Spray (2) 4- n, Drywell'and Torus Makeup Inlet (1)

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Drywell Makeup Inlet

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Torus Makeup Inlet (1) 3 s

p. let (1) , {

Nitrogen Purge Supply In l

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Dryve11 Floor Drain

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Drywell Equipment Drain

s. Treatment System (1) to Emergency Gas Vent

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Radiation Monitor (5).

Containment Atmosphere .

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) above list Reactor Water Sample (2 Excludwe from the v.-

- reviewed plant.

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.The above valves total'35 for the for.these y /

' he f ollowing: operating circuits l

>a: u The valve control Valves require the 4

dified to fore the isolation  !

a. Main valves Steam..isolationshould already condition existsbe mol switches to be placed verify that this Check to suggested modification. c

!l reset. not, the logic can be does If it on your plant. ented.

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....c-Figure 2-1 should be imp em va;.ves are controlled- ratort These and-require

c. hark All Motor-Operated , Valves " l" stritches ic has been re> 4pr b.

position spring-return-to valve Normaafter the isol ,

open the verify that this is true for a to A

to [

plant. as noted in 2.4 above.  ; !l closed-loop .<ystems on Those valves c.

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f .' bd 4. Those valves controlled.by switches which are spring loaded'to fw remain in the " Normal" or "Close" positiori as detailed in 2.3 -

gi . a, c,-and d above.) "

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,,l st1GGESTED' DESIGN MODITICATIONS

;ree design modifications are suggested in these guidelines. They are dis-'

. E' used below.

L-f a. -The first design modification involves tne Main Steam Isclation -

-s Valves (MSIV) which should be implemented if the design change has not already been implemented in your plant.

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s. V The second suggested modification involved the remaining solenoid / j f's pilot-operated valves for which a modification has been found to '

4 be necessary. This modification is detailed in Figure 2-2 and in ' '

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volves replacing the existing two-position, maintained-contact  !

'& switch'with a new three-position, spring-return-to " Normal" switch, -

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,, and the addition of two new relays for each valve circuit to be ..

$- modified. This modification is recommended for a permanent design a (

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change.- .It allows the trip logic to be res.et as soon as the cause '

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of the trtp clears without any additional operator' action, and al-3- lows the valves to be opened individually as required following  :

$ trip logic reset.

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'c. The third suggested modification is a more expedient temp -a-sign change. It is detailed in Figure 2-3. Thir. change *

'r putting " Closed" contacts from the existing maintained-- I U, two-position switches.in series with the reset relay fo Pri-y mary Containment Isolation System (PCIS) relay. This c!

may

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result in somewhat cumbersome operation. It does. howev acnieve-the desired result of preventing the reset action from e alng the isol utos valves by requiring each valve-operating avitch to be placed it; the "Close" position before reset can be achieved. Each  !

of the vi.1ves can then be opened individually as required follow--

) ing trip logic reset.

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.2.8 HARDWARE REQUIREMENTS l

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The following items of hardware may be used to implement the design changes -d y

suggested in these guidelines: * - I

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b,

a. Three Position Spring Return. Switches - These may be either CE '

if type SBM or CR2940 switches. The part number for each suitch is 'l b

shown below: .

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n1 GE Type SaH '

234A9337 P003 $

GE Type 29f.0 1~45C3040 P013 L~-

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'b .

W

_. v Relay - The suggested relay for these applications is the Agastat , e5 r e.1.ay . The appropriate relays for various operating voltages are . h.. .

shown below:

{P kv.

-Voltage Agastat Model r

GE Part Nr'her $

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120 VAC '

GPIC 750 164C5258-P001 T&,

I., 125 VDC GPDC 750 164C5258 P002 I+

.250 VDC GPFC 750 '164CS258 P003 I f r. .

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2.9 SAFETY CONSIDERATIONS

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The designs proposed in-these guidelines do not violate the IEEE-279 require-lments for separation of the various divisions. The. hardware suggested for use- i% ,

lare qualified types used in similar applications on requisition plants. Dur-

~ing ' implementation of these suggested designs, the BWR owner inust m ' . sure. *

. that all wiring and hardware locations meet - the IEEE-279 requirements for safe-ty systems. *  %

3, d 3 4 i

u.10 LPCl INJECTION VALVE LOGIC o E

It $

has been determined that with the RHR wystem In the shutdown coolinc

)

mode of operation, a low-watur-level . signal will automittically close the LPCI injection valves and a " scal-in" circuit will be energi::ed to hold a he valves

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f closed. This seal-in is in addition to the isolation logic " seal-in". A

- subseque it LPCI initiation signal vill not open the LPCI injection valves un- -

less the operator has manually reset both the logic r.nd the valve " seal-in" circuits. Failure of the operator to reset both t'.se logic and the valve cir- '

cuits will result in defeating LPCI initiation capability. Since the seal-in of the valve closing circuit is required to prevent the injection valve from i reopening when the suction valves close, it is suggested that operating d

g procedures be reviewed to verify that the plant is in the proper valve lineup h when coming out of the shutdown cooling mode.

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.g 2.11 OTHER CONCERNS WHICH MAY REQUIRE ACTION BY SOME PLANTS v ,

11 l f . n. Appendix A identifies a number of related concerns which have been re' viewed, c'),

and states a General Electric position for each. Each of these concerns 2 shouJ be reviewed and determined if they are applicable to your plant.

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} j-PRESENT V ALVE CONT ROL CtRCtRTS

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1 IP CHANNEL CtHCulTS '

DEStGNATIOst STEM vAtVE CONTROt C6HCtstTS ' PN90ARD OUTeOARO e

820 V AC ISUPPtY) V AttrE SOLCOfL 120 V AE 825 VDC 520 V AC125VOC b_ that VOt.T AGE A 8 ., C D "g.- - g4.* ISOL AT80N tht AY Kt

• ABCD ASCO ASCD

- yggp ASCO ,

TRsP REL AY K3

• A 3 5

- LOG 8C A A f RESET RELAY K4

• 2 2

JA , ,A 3 8 M ANUAL CONTROt.SW S

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2 REF ER TO CONTAtesMENT ISOLAT CH ANGE CONS 85TS S ASpCALLY OF:

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WLIA0E talTHESEPARAf RESET RELAY CONTACT tMG THE ASC ANO S&D A S C O g3 4 fclVALVES ADDING A NOtDfNG RELAY TO PREVEteT TO RESET DURtNG TESTigeG N HAV A _.

3. THESE CONT ACTS ARE CLOSEO IN THE **CLO 120 VAC (RETURN)  ;

8A- ITYPICA1. FOR TRfP CHANNELS 8.C.& DI -

• SEE TABLE AY RIGHT FOR COMPONENT DESIGNATIONS PRESENT CONFIGUR ATION

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. _ _ - - PROPOSED CONFIGUR ATtON PROPOSED V ALVE CONTROt. CtRCtetTS g 120 V Ac tSUPPLYi '

DEstGNATION eJ ITEM I OUT80ARD

% GN9QARD l l N VALVE SOL Cost K3 * ,

125 VDC 120 V AC 125 VOCl20 ML l LOG C* _j COIL VOLT AGE D

~~ KI . 1 A B C f x3.*

Ki * , Ki * ,*~ yt * ,

• " eSOLATION RELAY Kt

• B C i D K4

• A

• ~ 9~ HOLDtNG emMG ret AY K2

• A.C A. C S. O ret.AY K ~1-- S. O TRIP ret.AY K3

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K2 * ~ 1 MAleUAR. CONTROL SW S .* ~

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' RELAY PROPOSED T RtP CHAteNEt. CIRCtRTS ,

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Si 5-L Si Sct -

,_.c TRtP LOG 8C*

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~I " TRIP HEt.AY K3

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\ C D SEE NOTE 3 HOLOeNG RELAY K2

• S 3 A 8

• A A

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- 5 g) . RESET REL AY K4 MANUAt. CONTROL SW S

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'51L T Abt E AT HIGHT f 044 1N V AC (8tETUf tNI .

Obt.1hatJEt47 DEstGtJATIOts5 l'igtete 2-1. IISIV I. ogle Cluenge

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A1AfNI AINED I 'PCIS R5 LAY 2 1

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CONTACT ~~ HELAY ~~ '

SE AL-IN 5 SWs TCH gy - SE AL-IN IT '

, 2 MOMENTARY PCIS .=

• • 2 CONTACT- "

RELAY " "

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- RELAY 2 6 2T SWWITCH TRIP * '

I . LOGIC ,,

RELAY i

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' VALVE PCBS < 4 VALVE '

SOLENO40 MELAY- RELAY 2

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CONTACT CLOSE: AUTO /OPEN CONTACTS CLOSE NORMAL OPEN 1 - 1T X

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l NOTE: SY USING OTHER SWWITCH DEVEt OPMENTS.ONE RELAY

g. PER VALVE COULO BE EllwtNATED g SUGGESTED DEStGN L .

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l rigure-2-2. Recorm en.l.nl liesign Change i.er SolenoltI/I'llot Walvers e  %

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', bp I J._ tt.sIntained- I._ __ Mainta ined 37 ac of contact .j contact IT Switch 11 Suitch

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- Reset -- Relay  ;

2 Position. l Seal-in 3 Switches Seal in I 187 -

- PCIS a Pf:IS Relay 3 Relay-

~ ~ Reset Relay 25 I TRIP LOGIC eo '

TRIP ..

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,o LOCIC 4 g o .

• PCIS PCIS Valve g,g,y Valve Solenoid Relmy potenoiJ.

I Sog ested Design

! Present Iksign .

C0tiTACT CLOSE AlfTn/OPEN X ,

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Figure 2-3. Temporary Design Change for Solenoid / Pilot Valves ,

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3. DEFINITION OF " ESSENTIAL" AND "NON-ESSENTIAL" SYSTEMS .

,' PER NUREG-0578 REQUIREMENT 2.1.4 1

-1 Table 3-1 provides a generic assessment of those BWR systems which can be con-aldered " Essential" or "Non-Essential" for isolation conditions cons,1 stent with

,TREG-0578, Requirement 2.1.4. As used in this assessment, those systems iden-

ified as essential are regarded as indispensable' or are back-up systems in the

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',* avant of a loss-of-coolant accident. - The non-essential systems have been judged

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• - E' :o be not required in loss-of-coolant accident situations. However, dependir.g s

f? .pon the circumstances, it may be highly desirable not to isolate a "nor essen--

3-

< t gal", system. Fo'r this reason, and because the NUREG-0578 definition of "es-l

'i (y ;cncial" is deliberately flexible, the specification of " essential'.' is very ,

c.

L~, 1. - I4Jgmentalwithcertainsystems. Systems once isolated should be capable of.  !

i s-i Ning quickly returned to service as the need arises.

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[ lill SWR's have s,imilar safety system components. Tabla 3-1 is a' composite of LT./1-6 systems, and as such some systems may appear more than once because, de--

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anding' on the plant design, a safety system may be either a primary safety'sys "

. ;cm, or a sub-system of a larger saf ety system. Obviously, because the table is i eomposite, no specific plant would have all the' systems identified.

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Table 3-1~ '

ESSENTIAL / NONESSENTIAL EQUIPNENT a

System _ Essential Coussents

1. Reactor Head Cooling No Not a safety system-

2. Standby Liquid Control Yes should be available as back-up to CRD synstem

3. Isolation Condenser Yes High pressure safety system

4. . Low Pressure Coolant Injection Ye6 Safety system i

~'

5. -High Pressure Coolant Injection Yes Safety system i i

6. Core Spray (High-Low Pressure) Yes Ssfety isy' stem i

7. Closed Cooling Water ~ No Used for norms 1 operation only. Not, required'for DBA '

but is necessary for the recire, cleanup syst.we oper- ,

ation, and fuel pool heat exchsogers.  !

Y

" 8. Containment Atmospheric Control Yes Combustible gas control function necessary'. to e limin- Y '

- ate hydrogen / oxygen combustible at -_uf.;re. ' on

9. Dryve11/ Containment' Spray Cooling Yes - n Necessary to control drywell/contalement pressure

10. Automatic Depressurization System Yes Safety system; control of RPV pressure 11.

Standby Cas Treatment Yes Necessary to control emissions to envirousent ' .i

12. Reactor Building. Emergency Cooling Yes

]

Necessary to cool safety system pumps and motors

13. Reactor Core Isolation Cooling. Yes .Necessary for core cooldown'following isolation from  !

the turbine condenser and feeduster askeup. i

14. -Reactor Building Equipment' Drain Yes/No.

If drain :is -required.: the equipment 'is probebly out-

. of-service; check . for independent isolation; drain 1

i should,not.back up and flood essottal equipment.

15. Drywell Floor. Drain No -Not necessary for core cooldoen >

16. Emergency Service Water System Yes Necessary to remove' heat following. accident.. In-

cludes the ultiaste heat sink. .

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, - . - - - ,_ . _ . _ _ - _ _ _ _ - = _ _ _ - _ _ _ _ _ _ . _ _ _ - _ - _- . _ _

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ESSENTIAL / NONESSENTIAL EQUIPMENT (continued)- H System . Essential t'-te - ,

17. Instrument Air Yes Regarded as essential because this system supports esfety  ;

equipment.' Back-up accumulators are available for the-safety equipment should.the system fail.

18 ~. Service Air No ~ Serves no safety or shutdown. function.

19. Main Steam Line No Not required for shutdown.

p 20. Feedwater Line No Not required for shutdown. Portion that is Class I is  ;

j essential.  !

^ ~

I

21. Reactor Water Sample No Not required for shutdoun, but uould'be necessary for _

post-acc$ dent assessment.- Foot-accident osmple is'a l separate issue..  ;

u 22. Control Rod Drive Cooling Yes No credit taken for reflood, but-is. desirable.  !

,T Reactor Water' Cleanup ~

23, No Not required during and immediately following an acci-dent. Necessary in long-term recovery. .

Ih

=

! 24. Radwaste Collection No Not required for shutdown. ,

25. Recirculation' System No Not required for jet pump plants because' core can be cooled by natural circulation. EMR/2's must ensure that 't a minimum number of pump ' suction and discharge valves are open.? .

~i l 26. RHR Neat Exchangers Yes Main heat' sink during isolation.  !

l

27. RHR Shutdown Cooling. No Not essential but desirable to use if available. Not ,

redundant, but safety grade.  ;

28. RHR Vessel Need Spray No Not safety system.

29. RHR Drywell/ Suppression Pool' Spray Yes Necessary to control pressure.-

30. R'H!*. - LPCI Function Yes Safety function..

31. RHR.- Steam Condensing Function No .Not required as safety equipment.

}

32. Waste Collector'and Surge Tank'- No Not-required for shutdown. _

?

, . . , . . . . _ . . . ~ . _ .- - , .

. ,m , _ _ , . _ _ _ _ _ _ _ _ _ _ _, _ _ _ _ . _ _ _ _ _ _ . _ _ _ . _ _ _ _ _ _ , __.;..,,.,_,___,.

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Table 3-1 ESSENTIAL / NONESSENTIAL EQUIPMENT (continued) .

Systen Essential- ~!

r== ants.

33. Drywell Cooling -No  !

Used only in normal operation.- Desirable to-keep g running..

34. Demineralized Water No Not assumed available in~ECCS analysis.

35. Condensate Water - Yes/No Not assumed available in ECCS analysis, but is used '

in isolation condenser secondary.

36. -Fuel Pool Cooling No Boiling is acceptable, but aske-up is necessary. -

i Neat exchangers cooled by RECCW system.

37. Drywell Bleed - -Yes . '

Pressure control vent. Back-up to hydrogen control. .

, 38. Positive Seal System Yes l

Y Insures that highly. radioactive fluids are confined 'i

, a. . to the reactor building._ it, -

39.

Traversing In-Core Probe (TIP) No.

Not required for reactor shutdoun cooling. M

40. u.

Fire Protection Yes

Availability ~is essential, as the_" accident" may be

'

the result of a fire. Auxiliary askeup for isolation condenser secondary.. ~

41. Make-up Water Trestment No Serves no purpose during and immediately after acci- +

dent. Imager-term availability necessary.

42. -Feedwater Coolant Injection q

Yes - 'Like NPCI.- .

!. l 8

43. Separate Suppression Pool' Cooling Yes Main heat sink during isolation. ~ -r I

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g - ,.~- __ ,,.y., 7a, _. x -..mm n-g. -# -w-- -- g , sy - sw -3 g >- -

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?31 4. SYSTEMS CONTAINING RADI0 ACTIVE FLUIDS

'j ; PER NUREG-0578 REQUIREhENT 2.1.6a .

4-A' review of'the'composit'a BWR systems wa&' performed.to determine which systems

.!s should be regarded as containing highly radioactive fluids in the event of an.

j . .

. accident. Tables 4-1 and 4-2 provide 11sts of those systems which should and

- $: should not be regarded as containing highly radioactive fluid.

-1 All BWRs have similar safety system components. The. tables contain'a composite-( of BWR/1-6 systems, and as such some systems may appear more than once because,;

pepending on the plant design, a safety system may be either a primary safety syst'em, or a sub-syst'em of a larger safety system. Obviously, because the

-j-

• table is a composite, no specific plant would have all the systems identified.

I^

a I 1

4 '

. 1 1

1 1 l

g. I 1-

, l

.; j .- 1 q

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1 4-1 l, -

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, NEDO-24782

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Table.4-1 SYST*ES CONTAINING HIGHLY RADI0 ACTIVE FLUIDS

• eystsa Comments

1. Isolation Condenser Tube side only 1

'1 -

'2. Lov Pressure Coolant Injecti'on l' .3. High Pressure Coolant Injection

.;- 4. Core Spray (Bigh/ Low Pressure)-

1

5. Containment' Atmospheric Control'

6. .Drywell/ Containment Spray '.

L l 7. Automatic Depressurisation System'

/ '

Piping containing primary,:oolant only. Does not 1nclude instraent

. v hir and accumulators.

l- 8. Standby Gas Treatment L

l. 9. Reactor Building Energency Cooling Assumes primary coolant leakage in;

. equipment rooms.

10. Reactor Core Isolation Cooling

11. Reactor Building Equipment Drain I ~

12. .Drywell Floor Drain

13. Reactor Water Sample '

s c 14 '. Reactor Water Cleanup L i I- 15. Radweste Collectioni-u >

JJ vw

.16. = Recirculaeion

'. 17. RER He'at Exhangers Primary side f, I

-18; -UlR Shutdown Cooling Primary side h

19. RHR' Vessel. Head Spray I

, 20. RHR Drywell/ Suppression Pool Primary side Cooling :i

21. RHR LPCI Function

22. RHR Staan Condensing Function Primary side (if used during an accident).

23. Waste Collector and Surge Tank I

i 24. Fuel Pool Cooling .

25. Dryvell/ Containment-Bleed Assumed used during/afeer the accident.

26. Positive Seal Only on suction-type systems.

't 4-2

l. ,

1

. _ _ . - . . . ,_, _ . . . _ .. . . . _ . . . . . . . . . . . . - . . . _ .m. _ . _ . . ,

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'41 Table 4-2 -

t 3 SYSTDfS NOT CONTAINING RADI0 ACTIVE FLUIDS

. g. .

- t! Systen . Comments '

1. Reactor Head Cooling The Reactor Head Cooling System

' utilises water from the condensate' A'

storage tank and therefore should not contain radioactive fluids, g 2. Standby Liquid Control This system injects fluid and does- I not circulate primary coolant.. <

r

-. 3. Isolation Condenser Not in contact-with' radioactive (Shell Side) .

fluida unless tubes have failed.

1 4.' Closed Cooling Water This system does not circulate 1

primary coolant and could.only ,

_.. ./ become contaminated due to system leaks.

1 ( 5. Service Water This system does not circulate' l primary coolant and'could only

- a

'~ ,

become contaminated due to system-leaks.

.. 6. Inscrument Air Would require system failure.

7. Service Air- Would require system failure.

t 8. Main Steam Line Would require failure of isolation - -. t valves (and failure of positive seal system for those plants which havet this system). '

1 .- 9. Feedwater Line Same as for Item 8.

10. Control' Rod Drive Cooling Flows into pressure vessel. ' Water source is fros. condensate storage ,

tank.

I

11. Drywell. Coolers Have isolated on high drywell .

pressure for LOCA. '

12. Domineralized' Water

13. Condensate Water only if isolation valve leaks.

. De' pendent on' BOP ~ design.

14. Fire Protection System In-flow to containment.

15. Make-up Water Treatment i t

4-3/4-4 S th $' , .

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q , '..

t

i. '.3: IED06-24/82 0

o

-0 g' 5. EQUIPMENT REQUIRING ACCESS FOLINING AN ACCIDINT

- p, PER NUREG-0578 REQUIREMENT 2.1.6B The systems and equipment which any require access after an accident without .

considering equipment malfunction or any cleanup effort are: -

6 .

1. Outside the' reactor building .

Control room / building -

Radweste

, ivitchgear 2.

'Inside the reactor building or in a potentially high radiation area:

. v H2 and 02 analysers.

Remote shutdown panel Reactor water sample panel RPS. instrument racks 3

• Drywell atmosphere compressors or N2. supply Standby gas teatment systs:3. (filter change if necessary)

Reacter built; g vent or stack radiation Monitor-rack. 1-

'RHR/ service water spool piece This list is intended to be as generic as possible; plant-unique confirmation should be done and modification as necessary should be made.

9 O

5-1/5-2

.:.- NEDO-24782

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L ~ ,. .

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~

- 6. RADIATION SOURCE TERM INFORMATION .

PER NUREG-0578 REQUIREMENT 2.1.6B ,

! Appendix B ' contains the results of an analysis.to provide source ters'infor-' i

~

y estion for use in implementing NUREG-0578 Requirement 2.1.65, " Design Review - .

' of Plant Shielding 'of Spaces for Post Accident Operations." This computer out-

'put is based on an assumed operating history of 1 MWt for 3 years at which time 100% of-the noble gases 50% of the halogens and 1% of the remaining fission products are released from the core.

The user should multiply this computer output by the rated reactor power level and ' divide by the volume of water in' the suppression pool post-LOCA to determine ,

they activity per unit volume in the suppression pool liquid. It should be-emphasized that only the halogens, volatile solids, and non-volatile solids p should be so treated. Since the noble gases are net assumed to be absorbed in.thesuppressionpool,perNRCReguistoryGuide1.7,theyshouldnotbe i considered in the evaluation of the suppression pool activity post-LOCA.

This information will not vary app'reciably with the use of extended fuel-cycles.

l l

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. 6-1/6-2

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. WEDO-24782 4

"" . 7.

BWR GENERIC SAMPLING SYSTEM CONCEPTUAL DESIGN

.PER NUREG-0578 REQUIREMENT 2.1.8.A I '

7.1 INTRODU; TION This section presents a cone.eptual design for equipment additions needed to meet the Post-Accident Sampling requirem'nts e of NUREG-0578. This description was required by NUREG-0578 to be available by January 5,1980. The design 1s-preliminary in the sense that its' details are subject to discussion and refine-ment for individual preferences and plant differences.

Thja conceptual design described was prepared specifically for the Pilgria 1 Nuclear Power Station. Most of the concepts are, however, applicable to all geratingBWRs. It will be, of course, necessary to define plant-specific locations for the equipment.

The design is intended to'minimise complexity and maintenance needs. The

' result is that a three-step sampling process is recommended. The samples are obtained via a Post-Adcident Sample Station to be located adjacent to the secondary containment building such that_the sample room becomes an extension of_the secondary containment. This area will be properly shielded from the high dose rates present;in the secondary containment following an.

LOCA. The samples are then transported to a preparation area which consists of a simple wet lab with capability to prepare the samples for cou.; ting.= The final step in the process employs a counting area to be located far enough away from the radiation sources for accurate counting.

General Electric suggests a joint owners review of this Post-Accident Sampling _

Conceptual Design and others which are-in preparation by individual utilities.

Such a reivew might permit more uniformity.of implementation than currently exists, and also might permit joint design effort for the sample station' portion-of the plant modification. -

l 7-1 t

._ a ..- ., - - _ - - .~. _. - - . _ . - - _ _.. .

.c NEDO-24782 I

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7.2 DESIGN OBJECTIVES '

-7.2.1 Scope

.To obtain representative liquid and gas samples from within the primary con- .

tainment for radiological analys'is' in association with the possible event of ~

{

a loss-of-coolant accident (LOCA). The liquid samples will be. representative' ,

of liquids within the reactor pressure vessel-and suppression pool areas. Th t gas samples will be representative of'the atmosphere within the drywell'and- '

watwell areas.

/

The samples will bet w...,

s. Collected in a sample room.

'b. Sent to a wet analysis / sample preparation-lab. '

~l

c. Sent to a counting lab.

4 The post-LOCA sampling system will be designed with personnel safety as'its main feature. i 7.2.2 Criteria .

a. Liquid samples will be taken from two sample locations from within the reactor pressure vessel and two from the suppression pool. -Gas '

samples will be taken from two locations in the dryvell and two in j the wetwell areas. l

b. D A dose rates to personnel in the sample room wi'l be kept to at.ctoximately 0.1 rem whole body and 0.5 rea' extremities.

c. Sampling must not interfere with normal plant operations, s

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's, d. Design of the sampling system must be such.that samples can be.

. ]

gathered at any time during normal plant operation or during _ . 1 possible accident and transient conditions; i.e., liquid samples taken'at any pressure from 1400*psis through 0.psis, gas samples .)

.\

at 100 pois through -2 pois. * '

g .

e._ Liquid-and gas sample purge flow will be returned to the torus.

I i

f. Any sample line isolation valves must have an override on any auto-  ;

matic isolation signal to them. The ability to override must be- y e

, available from both the control room and sample room areas. ,

g. _ Liquid sample lines shall be classified and designed to Seismic Class I up to and including the second isolation valve. Beyond the  ;

second isolation valve the system will be classified .but not limited  :

to, Quality Group D, Non-seismic.-

1

h. Reactor pressure or normally operating sample pumps-will be used.to transfer-liquidsamples'throughthesamplepipyg. This '.' normal" motive, force will be supplemented using canned or bellows-type  ;

pumps. t .

3

1. Liquid sample. lines should be 1/2-inch achedule.80 minimum,IType-316L '

stainless steel with 0.020% C max 4um. The lines will be butt: welded with bent turns. Socketed fittings will be kept to a minimum. Tub- I ing can be used-to route _the sample around the sampler-panel. Tubing sizes can vary - depending on the application, to keep sample trans-: '

port time to less than 10 minutes, but tubing material requirements must stay the same. .

j. Cac sample. lines should be 1/2-inch minimum heavy-wall tubing,* stain-less steel Type-316L stainless steel, with high-pressure stainless steel fittings. Tubing sizes can vary when routing sacple'through the sampler panel, however, material requirements for this tubing must stay the same.

4 7-3  ;

. .g ..

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A separate Beating, Ventilating and Air Conditioning-(HVAC) Systec,

• j s

l will be required for air treatment within the sample roon. . Exhfus' '

i '

must be treated-and vented to the normal building vent system.

' Assuming it . takes 13 min'u'tes 'to obtain and prepare the samples.: th i- ,

' HVAC system design should be able to process the sample air volume 4 once during' this time. '

.s

1. The semple room, wet analysis / sample preparation room, and countin-

' room locations shall be outs,ide the shielding wall'of the secondar;.

containment, or reactor building. ,

t

' 7.3 ANALYTICAL REQUIRENDITS i

., - V  ;

7 . 3.1' NUREG-0578 Requirements Per NUREG-0578 and Harold R. Denton's letter, October 30, 1979 Section 2.1.( ;

" Improved Post-Accident Sampling Capability " the capability of' performing-l analysis on liquid samples ranging from 0.1 pCi/gr. (normal operating range)4 10.0Ci/gr.(RegulatchyGuide1.3 conditions)isrequired.

7.3.1.1' Liquid 4 -

~

a.' Radiciogical spectrum analysis

b. Dissolved gases, hydrogen, and oxygen

c. Boron ,

e

d. Chlorides l

l L -

e. pH -

[h 7-4 I*

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,yk, . , 7.3.1.2--Gas e S, a.. Radiological spectrum analysis

.- T .

. b. Bydrogen -

7.3.2 Internretation

• The following interpretation of the-NUREC-0578 requirements is applicable to the BWR.

/

J,3.2.1 Liquid g,

s =

a.. Radiological spectrum analysis

( ,

. . b. Dissolved gases, hydrogen, and oxygen

c. In-line-conductivity 4-td. An exception is taken to the measurement of bo'ron because it is

>t i not a normal post-LOCA ingredient in a Boiling Water Reactor ,(BWR).

4 c's. . An exception is taken to the seasurement of chlorides.under-LOCA

( " conditions. Chlorides are not a normal post-accident ingredient-in a BWR. The in-line conductivity measurement.taken in the sample.

room can be used to determine the range of chloride concentrations.

+ li

]- f.

An exception is taken to the measurement of pH under 7,0CA conditions.

A'BWR does not use chmaical control. The in-line conductivity meter j installed in the sample room will give an upper limit on hydrogen

j. .(H2 ) i " * ""*****"i "*

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7,.3.2.2 ' Gas . , . ,

y .

a. -Radiological spectrum analysis x o .

8

, b. Dissjlved gases, hydrogen, and oxygen t

e

. .- 7.3.3 Sensitivity of Analytical Procedures-l W ).,

s.

+

The NRC has not defined the sensitivity. The following recommenda-tions are reasonable requirements for analytical detection limit

/

/ a. Hydrogen (H 7

) or Oxygen (02 ): 1.0% (atmospheric-s.amples)

A v' (mole fraction).

b. Dissolved gases: > ambient temperature and 1 atmosphere

'solubili:y liair..

_ c. Individual nuclides

10% of the total samma. activity.

7.4. DESIGN CONCEPT i

7.4.1 Reactor Pressure Vessel (RPV) Liquid Samples.

When considering water samples taken from within the RPV, the following locations were reviewed.

7.4.1.1 Botton Head Drain This line enters the RPV at the bottom-most or "0" level designation. A sample tap l could be connected to this piping at some accessible location within the drywell. The sample from this point would be representative of RPV water. However, its location is prone to' blockage. Depending on where the sample tap could be located, considerable sample line purging would be required before a true RPV sample was attained. A suitable primary contain-ment penetration would be required to get the sample out.

l i

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.7-6 T . . . , ,,-

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- + NEDO-24782 l

l j ,, :7.6.1.2 Standby Liquid Control Line/ Core Delta Pressure Line

.1 V . .

o This line is a combination line serving two distinct functions. It uses one ,,

RPV penetration by-utilising a pipe-within-a-pipe principle. They are positioned inside the shroud at a location which would sample RPV water. Out-

. side the RPV the single pipe-within-a-pipe is separated by a tee. A sample-tap would be installed on either line from this tee prior to the_ isolation valves located within the drywell. The core' delta pressure line terminates with an open-ended pipe extenains slightly above the core plate. During an-LOCA there is a possibility of damage to this opening, therefore making it a questionable sampling point. A suitable primary containment penetration would be required to get the sample out.

.4.i.3 Reactor Recirculation Suction Line A sample taken from this location would be questionable and probably not representative of the liquid within the core region due to line stagnation and its location outside the shroud. A suitable primary containment penetration

, would be required,to get the sample out.

7.4.1.4 Reactor Recirculation Discharge Line As with the suction line, a sample taken'from any point downstream of the pump discharge isolation valve would be questionable and probably not representative of the liquid within the core region due-to line stagnation and its location with respect to the core region.

7.4.1.5 Jet Pump Flow Sensing Instrument Lines in the,, plant on'which this study was based ,all twanty jet pumps are instrumented

=to sense flow during normal plant operation._ Of the twenty jet pumps, four-have two instrument lines attached. These are jet pumps Nos. 5,10,15 and 20.

The twenty sensing lines sense incoming pressure to each diffuser, whereas the second s'ensing line on four of the pumps sense exit flow pressure. The sensing t

T -

1 4 ,

7-7 '

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-t [ lines, twelve total, for each group of ten jet pumps are gathered- together ~ a:

[ exit the' RPV at approximately the 105' and 285* asinuths.- They then exit -

through penetrations I-40A, B, C, ' and D in the primary containment at which 1 there'is a root valve and high flow' check valve on each,line._ For sampling,'

lower sensing tap on jet pumps 5 and 15 should be considered. Samples from t location would be representative of the RPV water. These areas are also welf protected from any possible internal damage.

9 7.4.1.6 Residual Heat Removal (RNR)' System-

~

The sample.systen could use the driving force from the RER system pump dische f when aligned in the shutdown cooling mode of system operation. The system ta j RPV water from a point ,in a reactor recirculation suction line.

7.4.1.7.' Conclusions (Refer to Figures 7-1 and 7-2)

, The jet pump flow sensing insenssentation lines located 'at the lower sensing-tap locations of jet pumps 5 and,15 would be the reconsended choice to obtain water samples from within the RPV when the RPV is pressurized; when-the RPV i depressurized an additional sample pumping system will be needed to transpor?

~

a sample to the sample room. Therefore, the RHR system should be considered' for low or depressurized $tPV conditions for the active force required to obts an RPV water sample.

Sample lines could be tied into the jet pump flow sensing instrumentation lin-at a point where they penetrate the primary containment, penetrations X-40A, :

C, and D. The sample tap would be installed at a-point between the root valvi and the check valve coming ~ from jet pumps 5 and 15 lower instrument lines.

From this tap, a' manual isolation valve would be installed from which-the saa-line could be run.to the sample room. Thees runs of piping must be adequate 1; chielded and protected in areas where operations personnel any be working.

4' The low-pressure sample taps would'be. installed on a discharge side of the RHR pumps, downstream of its backflow check valve. A weld-o-let could easily._

be installed in this section of RHR piping from which the sample line could be run to the sample room. A manual isolation valve will be required at a po:

nur the sample tap location.

7-8 br _ ._ . - - - - . _ - _

'. / ,. NED0-24782-

.f 7.4.2 Suppression Pool Liquid Samples 1

. :H -

Sample points from the following penetrations were considered.

3 t- 7.4.2.1 Level Taps (X-206B and C) ^

't- ,

.i

~; Line size is 1 inch and lines are used as liquid level indicators. These 7 i . 1 g lines are the lower or liquid leg lines. The location and size of these lines 4 make them desirable for sampling suppression pool water. Crud accumulation  !

.' would be minimized.

3

/

l' y- 7.4.2'.2 Drains (X-213A and B) j- '#

p'  ;

• Line size is 8 inches and lines are used as flanged construction drains. To q j put in a sample point on.either of,these connections, the suppression pool' p would have to be drained; Their location, the bottom-most area of the suppres-

. sion pool, would n'ake any s uple lines run from these points susceptible to damage and possibly contaminated with. accumulated crud.

v  :

i

I 7.4.2.3 RCIC and ECOS Suction (X-220, X-221, X-222, X-229) >

0 i e

i

[ . The X-220 line size is 6 inches and line is used as RCIC Pump Suction; X-221, ,

line size 16 inches is used as EPCI pump suction; X-222A, B,. C, and D, line -

size 18 inches are used as RHR pump suctin X229A and L line' size 18 inches 1 g

_l are used.as core spray pump suction. Sample points for these lines would ,

be on'any line downstream of the first isolation valve, but before the pump. '

suction for that particular system. Thir location would be susceptible to ,

v

, pump isolation and would possibly require override capability to.that isolation valve in order to obtain:a aample. A second sample point choice for any of

j. I c these lines wocid be.at a location in the line from where it exits the suppression pool to the'first isolation valve. -These locations would'be prone -t to crud accumulation and, considerin the pipe sizes, it would be necessary i to drain the suppression pool water below the line to install a sample tap. 3 6

7-9 LM__- - - - - - -

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,? NED0-24782 J .

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l . 1 7.4.2.4~ Residual Beat Removal (Rk./ System -

)

This system could be used for obtaining a suppression pool water sample when

]

the system is running in either suppression pool cooling or suppression pool to suppression pool test mode. ,

7.4.2.5. Conclusione (Refer to Figure No. 7-3) t

, penetration X2068 and C (level taps) would be a recossended choice for two su pression pool sample points. These two' locations along with' sample. taps plac

, on the diac'harge side of the RER pumps would give total sampling ability of t' .c

, t

' suppression pool water. The samples from the two instrument, sample taps woul

~~

be transported to the sample room via a separate-pumping system. This system would be used if the RRR system was not operating. The sample line for this:  ;

. system could tie, int'o that sample line coming from the RER pump discharge',~

downstream of the backflow check valva (the same sample tap location describe j J tor the RPV sample. This line, however, should be run separately from the line used for RPV samples.) This sample line would then be runto the sample  ;

room. These runs of piping must be adequately shielded and protected in areas where operations personnel may be working.

i 7.4.3 Drywell and Wetwell Gas-Samples The recounsendation is to utilise those 02 analyser lines which are not in use ';

The two drywell samples should come from two-elevations, preferably a locatic near the mid-poir; , reference. penetration X-50A, and a location in the upper-most region, reference penetration X-29D. The wetwell locations, reference i

. penetrations X-228C and J, would be.recoaNaded. Kcte that the penetration

-locations referred to-are for location requirements only. The physical saap3- -

points would be l'cated o outside the primary containment, beyond the isolation control valves but before the 02 analyser Panel, C-41. The isolation control valves on the 02 sample lines chosen to be used for the new gas sample syster e

O L 8

7-10 e .

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NED0-24782- l 4

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must have their operational c'ircuitry redesigned to include an override of the. j i

li isolation signal which would be controllable from both the main control room analyser sample lines'

~

^

and proposed sample room. By tapping into the present 02 l prior to the analyser panel-the system would require all associated hardware to transport the gas sample to the proposed' sample room. This would be a self-r contained and fully controllable system. An alternate, or back-up, system ,

could be one that utilises the presently installed ' sample pumps,- P-166A and 3, In either case, the and taking a sample off the discharge from these pumps.

serpies would be returned to the torus via a newly run return line using a- ,

analyser spara penetration such as 1-2887, or by utilizing the present 0,,

j return line to the torus.

7 In either case,_the gas sasrple lines, from the source of the sample tap'to tho' .

Noposedsamp1eroom,willrequireheattracingtoapproximately280'Fforcon-

~

~

densate prevention. This heat tracing should be turned on only af ter an 10CA-event. These runs of tubing must bt adequately shielded and protected where t

operations personnel any be working.

i i p

i 7.4.4 Proposed Sample Room Considerations- (Refer to Figures 7-4 through 7-9)

L .

l 7.4.4.1 Requirements (Minimus)

a. Personnel safety Ease of entry and exit b.

c. Must have its own communications system to main control room

d. Must have~a radi'acion monitoring system- ,

\.

Must have a separate power source or tied into the emergency diesel-e.

generating system Ventilation system capable of room volume air change every 15

.f .

minutes.

4 1 7-11

.i

.- _ . -_ __._____.___.___.__.____.____.____________.____________._____.__________m

WED0-24781

,.,I 0 i. . *

' Must be as close as possible to sample sources must be able to meet g.

I WUREG-0578 allotted sample time requirements

\ -

', b. het h, "ca own clean wetet source

i. Mast have its own bottled compressed air, or nitrogen supply system-

j. Must have the ability to handle and perform possible chemistry, dilutions, etc., on any samples coming into the roon

• c , k. Must have its own drain system, pumped to the suppression pool.

~- ' 7.4.4.2 Location Considerations (Based on Pilgria Site)

a. The present Health Physics office location would be acceptable.

Additional shielding would be required along the Equipment Removal Hatch wall, Airlock wall, and secondary Containment Access Lock

' wall. shielding requirements need to be assessed. Other require-ments such ad 1 hose previously mentioned wdd have to be investigated to determine if this area is adequate.

b. A' secondary structure built as an addition to the Hea1th Physics

' office or a separate addition to the building in this.same area.

. Shielding requirements for this area require assessment. Note locations are shown on Drawing M-16, ground elevation, 23'-0",

between N.1 and M.5 and Line 5 west.

c. A location 6hown on i)rsving No. M-18. El. 51'-0" in the fan room side of secondary containment "J" line, between 11 and 17. This area is in close proximity to the same point locations and building ventilation systems and has ease of entry and exit by way of the turbine floor. Dose rates in this area 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />' into an LOCA are conservatively estf u ted to be 1-3 Rea/h (see dose rate calcula-tions in Figure 7-10). The fencing along the east-west line, parallel to the generator will have to be moved to a north-soutn 4

7-12

l

. *.:;r

- 1 I

WED0-24732

, . - , l I

.. j

\ ' position along the alternator and of the generator, giving personnel access to the turbine building truck hatch. A spiral staircase, or -

, other mode of access, will be required for personnel to get from the 51' elevation to the ground, al'evation 23'. Some type of mechanical device could be designed which would transpor:.the sample between these same elevations, thereby'giving personnel more freedom when l

\

descending the stairs. ,

d.

Conclusions:

1 The location described in c. above would be roccesended for the plant

/ on which this stud;;< was based. Locations are expected to be very

. .,- plant-specific.

7.4.4.2 Sampler Requirements .

a. Dissolved Gas and 1.arge-Voluise Liquid Sampler (Refer to Figure 7-6)

The dissolved gas and large-volume liquid sampler will be designed to take large m lume samples as well as measure the volua,e of dissolved gases in a pressurised liquid and to obtain a sample of

. s these gases for activity and gas chronstography analysis. Large-

volume liquid samples may be required for conditions less severe

, than a Design Basis Accident.  ;

l

. For tho SWR it n recommended that the dissolved gas composition l t er.d cur. centration normally be determined by sampling and analysis l '

e of the $as pha/se over any liquid volume and application of Henry's l Law togsthat with esasureatut of.the applied pressure. This type of I sampling system is required if a scenario can be hypothesized that results in a gas bubble ir the pressure vessel with the primary system isolated an# pressurized. AnM ysis 9f the composition of i the gas phase is then determined by applying Weary's Lau to the

. dissolved gassa. It may be possible eith furtPer analysts to eliminate the expansion buret n.d . hypodermic needle. Because of the largo quantities of potentia. 6as activi'rf es it is recommended 1

that the sealed venting system ht retu.ined for degassing the sample. -

l 7-13

j WEDO-24732

. i The recirculation pump serves to break down supersaturation of th dissolved gases by agitation after initial depressurization into espansion buret. It should be a magnetically coupled, saaled.

centrifugal type.

A standard 14 milliliter offgas septum vial is evacuated on the atmospheric saapier and transferred to this hypodermic needle. A autended tool is required for insertion and removal of the vial.

In-line sample monitoring for condictivity and radiation will be j included as part of the sampling system.

- 4

b. Small-Volume Liquid Samples (Refer to Figure 7-7)

The smal)-volume liquid sampler is designed to take a defined voi of sample, in this case 0.1 milliliter, over the complete range o possible system pressures. The sampler arrangement is shown in Figure 7-7. Basically, it consists of a four-way ball or pis.g vs where the volume of the sample is defined by the bore of the valv A syringe containing a measured amount of diluent is attached to h upper Luer-Lok fitting prior to sampling. The septum-type sampli bottle is also mounted on the hypodermic needles prior to samplir A purge flow is established through the sampling system and the c-stress valve is throttled to maintain the pressure as desired to minimise phase separation in the sample line. A by-pass line is provided in order to increase the purge flow, and consideration should be given to the necessity of providing a sample cooler. i a time sufficient to thoroughly flush the system, the downstresa valve is closed, the sample is isoleted, and the sampler valve is opened and the diluent in the syringe is injected through the sat valve into the sample bottle. If it is desired to also collect t dissolved gases, the procedure can be modified to use a pre-evact sample bottle and to keep the vent valve closed when injecting ti diluent. (The vent valve is included as a precautionary method <

y.

7-14

" l l

NED0-24782 l l

I f * '

[.

eliminating overpressure in the event sample bottle vacuum is lost E I during mounting of the bottle.)  ;

l l' l .

An accurate measure of the amount of d*1uent added esa be obtained by weighing the sample before and after samp1~1ng. Basdling tools j should be provided for mounting and removing the sample bottle and for inserting it into a shielded sample carrier.

c. Atmospheric Sampler (Refer to Figure 7-9) s This system would act normally be used to obtain todine and part-  !

?

iculate samples from the drywell.  ;

l -. v' l

' It is recommended that 47-am-diameter glass fiber filters be used f for collection of particulate activity. It is further recommended' l

that the sample cartridges and filters be housed in a laboratory .

drying oven at 280'F to prevent moistur precipitation. Silver Zeolite cartridges are recommended for total todine activity col-lection. It is questionable if the usual iodire chemical species t

characterisation media can be used at this temperature.  ;

i ,

Consideration will be given to using a critical orifice in the cartridge dischasse line as an alteraative to the orifice flowmeter and sample pressure measurement. Sample flow rates should be on the order of 0.1 to 0.5 cfm with a 1 to 5 pois anticipated pressure drop across the filter cartridge combination.

t

/

The gas-phase samp'ler is basically similat to the standard BWR ,

offgas sampler and uses septum-type sample bottles. The hypo-dermic needle should be a standard Luer-1.ock type and an extended installation / removal tool should be provided. An extended installation / removal tool should also be provided for the gas  !

sample bottles. Recommended sample volumes are 14 milliliter >

7-15 ,

9 n , . . . . . , , . . ~ - . . .

e

.', NED0-24782 c ,

(standard,offgaa sample vial) and 100 milliliter. The counting systems should be precalibrated for these sample volunts. If the

.. 14 milliliter sample has too much activity it is a simple matter using a 0.1 or 1.0 milliliter gas syringe to transfer a measured aliquot to another partially evacuated 14 milliliter vial. cas syringes would be used to take fractional milliliter samples for R2 and 02 analyses by gas chromatography.

.. a P2 should be a compound gauge. ranging from - 14.7 pais to the naziwa comisinnent design pressure. . ,

J SV-1 through SV-4 are actuated by a multiple, rotary selector  !

~

• switch. Sg prevents operating SV-3 unless a sample bottle is mount  !

on the hypodermic needle. 82 defines the actual moment of samplint >

j A cooler and liquid trap are recoussended to minimise failure of ,

flapper valves in tr.a pellows pumps. Consideration:should be giver.

^

to installing the pumps with quick disconnects in order. to simplify replacement.

t f d. Sample Roos ,

The sample room will house all the necessary equipment controls ,

required to transfer a sample to.this location. Preliminary infor-nation such as conductivity and radiation level will also be made  !

available. The samples will be gathered at this location and the first dilution of-those samples requiring dilution will be made in t

the sample room. Those equipment items which will be considered radioactively " hot" will be shielded and operated remotely via reachrods or special tools designed for specific tasks.

l

=

The motive force to transfer a watar sample from the RPV will be. -

provided by t.he pressure within the vessel. When pressure within '

the RPV is too low to move a sample to the sample room, a sample can be taken from the RHR pump discharge when aligned in the shutdown cooling mode.

, 7-16

. - _ . . - __ ._ _ . __ _ _ . _ - _ _ _ . ~ . , . . _

i TI-  !

R j J .

NEbo-24732 i

1

.( The suppression pool sample will also be routed to the sample room .

using its own pumping system or the RHR system which is set up in ..

~

the suppression pool cooling or suppression pool to suppression pool l test

• mode. .' I i

Metal-bellows-type psesps will be used to transfer the drywell and .

.I vetwell s. a samples to the sample room. Separate and redundant pumps wi3 be used for samples gathered in each area.

]

Return lines for both the liquid and gas, samples will be directed

];

back to the torus area so no ascess accumulation of irradiated ,

e materials are held or stored in the sample room. I s

~ '

Clean water will be available to flush all liquid sample lines in i the sample room, thereby again minimising any radiation buildup in l this area. ,The same is true for the gas sample lines. A bottled ,

compressed' air or nitrogen system will be utilised for flushing out f accumulated radioactive gases. This same bottled gas system can be. I used to blow out a clogged liquid or gas sample line, thereby giving [

extra assurance of obtaining a sample. ,

The e' ntire sample system should be designed such that all samples vent through a closed vent system as they are depressurised to  !

minimise the evolution of noble gases into the sampling area -

atmosphere. Great care should be taken to choose valves having a s high degressof assurance against stem leakage. l N

A " breakaway"-type torque wrench should be provided for ose on reachrod-operated manual valves to. reduce the possib'.11ty of seat damage due to overtightening. '

n D

e y e

e 4

1 7-17 s

O

r _. __ . . _ _ _ -. .._.-. _ ___ _ _ - . _ _ _ . _ . . _ .

. . , . =

.m  ;

v .

NEDO-24782 1

• Transfer va1ves, either manually operated by use of reachrods or ,

motor- or air-operated, would be controlled from the sample room or

control room. Work areas within the shielded protection of the sample room consisting of cubicles and fuma hoods for initial sample dilution will be provided. '

4 Regulatory Guide 1.3 LOCA assumptions result in primary containment (l iodine activity equivalent to spproximately 5 R/h per milliliter at f 1 foot. Approximately 30% of this dose rate is due to 52-minute  ;

1-134, 25% due to 6.7-hour 1-135, and 25% due to 2.28-hour I-132. i The balance is due to 1-131 and 21-hour I-133. Depending upon fuel-

' i l , '

t l

' ' ' lasching conditions, the I-132 concentrations may be' supported by the 72-hour half-life of its $b-132 precursor. Because of these- _;

i very high dose rates, .the sampling system should be designed to take liquid samples as small as 0.1 milliliter. It is estimated that f

dilution factors as high as 1 x 10 6may be required for counting I

purposes, i.e., 0.1 milliliter - 25 milliliter - 0.1 mil-  ;

l liliter ----* 25 mil 1111ter - - 0.1 milliliter, is a dilution factor of 6.25 x 10'. Therefore, in order to minimise personnel exposure, it is recommended that the initial dilution be made in the sample room.

Shielded sample containers should be provided to tranefer this sample to the wet analysis / sample preparation lab area. Surface dose rates jL of no more than 100 mR/h are estimated for this first dilution volume L

using a lead container weighing approximately 35 pounds. This is l

I taking into consideration at least a 1-hour decay time before the first sample can be taken. .

e. Wet Analysd i/ Sample Preparation Lab ,

y <

s In the plant on which this study is based there is only one area being considered for this f acility, the Augmented Of f-Cas (A.O.G)

Building. The building is well ventilated and has facilities for wet sample analysis. For example, the room housing CP-003 and I'

' MCC-B-31 is usable and if needed, could easily be enlarged toward 7-18 .

---m-----_-m_-_______.____ - _ ____ _ _ _ _ _ . , _ _ _ _ _ _ _ . - , , - . , _ , , - - _ . . . - , - _ , , , - _ , , , _ . _ ,_,.,1 ,. . - , . , , . - , , , . . - - - - . --,,y..--.

NED0-24732 l

.I g' ..

e the centerline of the building. Shielding requirements would need assessment. Storage areas for the samples could be provided. - l Auxiliary power for the analysis equipment could be orovided for by l the same independent power source Yhich is located in the Guard Shack. Clean water and bottled compressed air or nitrogen systems could easily be provided.

Concept. The laboratory provided for the accident case should i

contain as a minimum, a fume hood with an internal 2-inch lead .

1, shield and a sample storage area. Analytical equipment should l e

include a' top-loading balance with a capacity of at least 200 grams and a sensitivity of 0.05 gram for making weight dilutions.(this  !

~'

method results in far less operator exposure than is required when l' adjusting volumetric flasks to a level mark). A simple gas chromato-graph should be available for atmospheric hydrogen, oxygen, and i nitrogen analysis. A pH meter should be provid.d ,tions with a simple ,

i l

conductivity meter (this combination might eliminate the need for ,

more involved analysis by demonstrating twquired sample purity). A vacuun pump should be available for evacuating sample bottles and ,

for filtering. It is recommended that all dilution stirring be done with magnetic stirrers and there should be stirring bars which do r.ot >

have to be recovered from the potentially highly radioactive solu-tions (this will greatly minimise the contamination problems).. There  ;

should be a supply of gas syringes for gas sample dilutions and for ,

chromatographic analysis. Luer-Lok fittings, valves and hypodermic needles should be available. Micropipettes with disposable tips will lc also be required for sample dilution. Consideration should be.given ,

to using plastic " glassware" on a throw-away basis to minimize equip-ment cleaning and general contamination requirements. The laboratory should be well stocked at all times with gloves, wipes, decontsmin- >

ation solutions, lab coats, shoe covers, air masks, plastic bags., and

. absorbent floor and bench covering materials.

t 7-19 4

=v .

i i

I

.. j I

NED0-24782 I j, -

.g - .

A

f. Countina Lab .. i l

in the plant on which this study, was based the area presently located _

in the basement of the guard shack is proposed. shielding require- j ments for the counting equipment will have to be assessed along with additional lab shielding for personnel protection. The guard shack has its own independent power source and ventilation systems. )

Concept. The counting facilities should be located in an area which j

, will not exceed 1 mR/h under accident conditions. The detector

/ should be shielded by a minimum of 4 inches o* 1ead. The detector

^

~ ~ -

• should be contained in a volume which can be isolated and purged t e

with bottled compressed air. as it is conceivable that the ambient atmospheric noble gas activity might be high enough to swamp the detector. A gamma spectroscopy system consisting of an intrinsic i type germanium detector (62 efficiency relative to a 3 x 3 inch ,

Na1 detector would be sufficient) and at lagst a 1000-channel multi-channel analyser should be supplied. The analyser should be a hardwired type and should have automatic peak search and identifi- -i cation capability and calibrated to at least two shelf positions, a '

I factor of 100-500 spart in sensitivity. An intrinsic type, rather I than a lithium drifted germanium detector, is specified for reli-ability reasons. Reliability of a liquid nitrogen source must be -

assured and the detector should be maintained in liquid nitrogen at  ;

all times as it takes several hours to cool down to operating temperature. A sodium iodide detector should be available as a backup. A shielded sample storage facility should be made available in the counting area (at least 2 inches of lead for shielding).and arrangements should be made for long-term storage, possib1) in the wet analysis / preparation lab. It will be difficult to handle L samples reading more than 1-2 mR/h since it is possible that the ambient noble gas activity may make it virtually impossible to use i

extended geometries.

i b e

i G

7-20

4 4 e d 4 h.e - 4

- +

NED0-24732

)

( 7.4.5 Effluent Particulate and Iodine Samplina l The limiting factors associated with sampling the main stack effluent during )

post-LOCA conditions are those of personnel safety in getting to the present .

I

, sample room locatita and returning to the proposed counting room.- The radi- -

ation " shine" from the reactor building and the stack plume size and speed and '

its direction relative to the present stack sample room 'are the limiting factors t which should be cassidered. The present consideration of personnel safety and l radiation levels inside the present sample room could be tolerated for sample I removal.

i .

It may'be necessary to consider a new sampling room location for a post-LOCA

~

condition only. This nov location would be in an area as far away as possible  !

from the reactor building and perpendicular to the main stack and the prevail-ing wind directions. Safe distances,can be obtained using that information available in Section 7.5. '. System design would be dependent upon the area chosen ,

i- for the new samplins' room. It can, however, be recommended at thfs time that l- sampler cartridges of Silver Eeolite be used for radiciodine instead of the routine charcoal cartridge samplers.

An alternate to the above which might be considered ie that air which is being discharged from the Stand-By Gas Treatment (SGTS) System. This system is close to the proposed sample room location. The sample linea can run from the SGTS system discharge to the atmospheric sampler (refer to Figure 7-9) to obtain ,

. the required effluent samplan during LOCA conditions.

7.5 PILGRIM POST-LOCA ON-SITE DOSE RATES ,

An analysis was undertaken to estimate on-s,ite dose rates outside the reactor building (secondary containment) af ter a postulated accident resulting in Regulatory Guide 1.3 releases of radioactivity to the primary containment.

At t=0, it was assumed that 100% of the core inventory of noble gas and 25% ,

of the core inventory of iodines were airborne in the primary containment.

l-Figure 7-11 illustrates the plant model and leakage paths.

1  :

. 7-21

'f .

i-NED0-24782 f i

i

.~

Three cases were esamined in order to bou'nd the dose rate contributions from [

k stac'k releases and reactor building shine. The assumptions for each case are  !

listed in Table 7-1. "

Case 1 'r'sulted e in the largest relea'see ~of radioactivity from the stack becaus  !

no credit for mixing in the secondary containment was assumed and was, there-  !

fore, uskd to determine the ground level dose from the plume. The whole body ,

dose rate in mRam/h as a function of distance downwind of the stack is pre-  ;

seated in Figure 7-12. The dose rates at distancas less than 100 meters are conservative estimates, since the comput r code used does not calculate doses  ;

at less than 100 meters. The corresponding inhalation doses are presented in- l

\

Figure 7-13.-

, _ . ,-. Case 2 was used to determine the maximum reactor building shine contribution i since no leakage from the secondary contairement was assumed.

~

Ccse 3 was run to determine the actual time of peak secondary containment activity in order to chaosa the point in time for the building shine 5 calculation.

B i

Figure 7-10 graphs dose rate eersus distance from the reactor pressure vessel >

centerline. These correspond *o the activity in the secondary containment i

from Case 2 at 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> post-LOCA, the time of peak activity. The building  ;

shine calculations war.e performed for points along traverses south of the

• building at elevations of 0 feet (ground) and 51 feet (elevation of the pro-

[

posed auxiliary building sample station). The effects of the air-ground interface and of attenuation by the structure south of the building were not considered; consequently, the calculated dose rates should be conservative. . I It should be stressed ihat these calculations were performed assumina a power of 1 MWt; therefore, all dose rates must be multiplied by rated thermal power to obtain full-power dose rates.

I 7-22 u-

-,w e.------. -,n.. a,_,n,-- , .

.. . 1 NED0-24782 l

I The dose. rate at the sample, station at the base of the stack can be estimated as follows:

Whole body dose (los)* = 2.7 x 1998** =

5500 mR/h Buildia's Shine (20da) = (2.0-1) x 1998 = 400 mR/h

. 5900 mR/h I

• closese point analysed '

• • rated thermal power for Pilgria The inhalation dose is negligible close to the stack as can be seen from Figure 7-13.

s .

Doses

• at other points on-site can be estimated in a similar manner using Figures 7-10, 7-12, and 7-13.

t I

I t

1 e

l

)

s

^

l i- -

k i

7-23 I'

f

NED0-24782 i

Table 7-1 ASSUMPTIONS FOR RADIOLOGICAL ANALYSES i-

)

CASE 1

.L L 2-l 1 1 1 i Power (MWt)

Drywell Le nage (%/ day): L1 0.5 '0.5 0.5 Leakage from Reactor Building (%/ day): L2 100. O. 100.

Mixng in Secondary containment (%) O. 100. 100. ,

SCTS filter Efficiency (%)

90. 90. 90. f

'

• Meteorology's ' Stability C. Windepeed 2 meters /sec Source Ters: 100% Noble Gas and 25% of Iodines in primary containment at t = 0.

f

• An arbitrary, reasonably averags meteorological co.idition. t i

i I

)

l 4

0 l

7-24

. IMD0-24782

/ *

\.

• PROPbSED New aAMPLE ,

TAP LOCATION axisTINo ExisTINo BALL SALL' y$'$ JrrPuMPs y$8$ JET PuMPts LOWER LOWER-PENETRATION PENETRATION PROM PRIMARY PROM PRIMARY 00NTAINMENT CONTAINMENT

/ PROPOSEO MANUAL

, y' ISOLATION VALVE (TYPICAL) 4 t

PROPOSED ISOLATION -

VALVES I MO I MO l d2 &la v v, SECONDARY CONTAINMENT

. SHIELD WALL -

C

• SAMPLEROOM SHIELD WALL .. ,

SAMPLE ROOM (TRANSPER EOulPMENT CUSICLE)

)I ,

TO TIE IN REFER TO FlouRE 74 rigure 7-1. Pressurized RPV 1.iquid S' ampler e4 7-25 cg-a

- - - - ~- _ ~ . . . . . _ . - . _ _ - _ . _ _ _ , , _ _ _

i IED0-24782 i

PROPOSED EFISTING N 'E .

I

' TAP BACKPLOW LOCATION

- gMISTING CHECK RHR VALVI " l pump (TYPICAL) -

EXISTING (TYPICALI

- '  ; y N

• g gg

1. EXISTING V ALVE PROM RPV " O POR PLOW CONTROL (TYPICALI MANUAL pppy .

190LATION VALVE (TYPICAL) ,

• i l

". et '

l MO l p /- -

mi2 r,

m 7

PROPOSED  ;

i SAMPLE TAP LOCATION 1

PROPOSED l ISOLATION

• VALVES I MO I I MO I  :

m12 mim Vm VM r.

SECONDARY m

CONTAINMENT W ALL SAMPLE ROOM -

SHIELD WALL SAMPLE ROOM l- ' (TRANSF ER EOUIPMENT 1f CUSICLEl TO TIE lN REFER TO PIGURE 74 j.

Figure 7-2. Low-Pressure or Depressurized RPV Liquid Sampler (Reactor Pressure Vessel to Reactor Pressure Vessel Moo of Operation) 7-26

_ _7 l

• e NEDO-24782 )

SAMPLE 'A'(TYPICAL POR SAMPLE Tl k .

SUPPRESSION P00L 5

. 1 EXISTING l SUPPRESSION POOL WATER WATER LEVEL Lgygt  ?

INETRUMENTATION (TYPICALI PROPOSED SAMPLE TRANSPER '

PUMP X J \

. PROPOSED

. v" q y MANUAL ISOLATION i j k VALVE I PROPOSED SACKPLOW

r ExtSTING MTEM .

CHECK VALVES

. Ym. ,,

i i  !

PROPOSED  !

MANUAL

!. ISOLATION EXISTING - -

l EXISTING SACKPLOW i RHR CHECK V ALVE i-pyg, [ MO l EXISTING t _

, LI2 -

l

. ym l MO l PROPOSED l MO l PROPOSED f S. AMPLE 'S' [ Ll2 LI2 SAMPLE AS *A'

77 r7 SECONDARY CONTAINMENT WALL m .

SAMPLE ROOM C SHIE LO WALL SAMPLE ROOM (TRANSFER EQUIPMENT TO TIE IN CUSICLEl REFEA TO FIGURE 74 -

71gure 7-3. Suppression Pool Sample Line 7-27 [

-t _

NB:DO-24782

. ,  %/b

= i.

Ak SMIELD WALL \

ee - - EEIRIRL F M

-4 l-

• 1

'm SMIELD WALL '

1 TRANSFER PENETRATION d I a d

BOUIPMENT 3LEEVE I

. . CUSICLE

g .

I e

• SAMPLE g j l i

V < .

s i i a l - I f

6 --. .J y' b_..J BMIELD WALL /

.- 7

• -- - =' E NTR ANCE POR '

MAINTENANCE OP SMIELDED TRANSF ER EQUIPMENT WORK EECONDARY

= 20 N AREA- CONTAINMENT WALL s

i l BMIELDED gong SHIELD gpgg WALL I .

SMIELCdD WORK AREA I

y .

c =12ft 7 Figure 7-4. Sample Room Layout Concept 7-28

. - -.. . . - - . . - . . . . . . . . . . . _ . . . . - - . . . - . . . ~ - . - . - . - - . - - . . . . . . - .. . - - . ~ . - -

i IEEDO-24782 -

l 1

"~' SUPPRESSION PRESBURIZED LOW PRESBURE

*\ / POOL SAMPL4

%/ SAMPLE:  % / AND CrEPREstuR12ED  !

REFER TO REFER TO SAMPL&: REFER TO PlOURE 7 2 *

, PlOURE73 FIGURE 71 .

: PRE 88URE SENSOR  ;

PRESOURE

-e agN80R *

• 4 3 f. *1 fISOLATION VALVE. f '

ISOLATION VALVE N N Y

• I

.. PLOW

"* REffRICTING ORIPICE

• gj j CLEAN WATER r, '"

.k J W .tiNO i VM

'A J I VM

, SACKPLOW i

,' CHECK VALVE j

t>

• i

. e .

j V3 V == AC POR PLOW CONTROL 4 L- i 1f *

^ PRESSURE SENSOR

TEMPERATURE SENSOR

PLOW INDICATOR IN.LINE CONOUCTIVl4 .

IN.LINE RADIAT80N MONITOR RETUalN m *LJ f TO TORUS ' V7 I

U l TO LlOUID SAMPLER i REFER TO PIOURES 74 and 7 7

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NEDO-24782

(' 8.

• SUPPLEMENTART INFORMATION ON REACTOR COOLANT SYSTEM VENTING PER WUREG-0578 IMPLD(ENTATION LETTER REQUIREMENT This section. addresses a number of issues in support of the adequacy of existing BWRsystemstofulfilltheintentoftheifbEEG-0578implementationletter .

requirement for reactor coolant systen venting.

8.1 BWR OWNER $' CROUP POSITION ISSUES The BWR Owners' Croup generic position issued October 17, 1979 (Appendix C) committed four actions for additional study in support of that position

,1. Whether it is desirable to interfere with emergency core cooling

~

' functions in order to prevent venting the reactor systemt

2. Provision of plant procedures to govern the use of the relief

~

valves for venting the reactor pressure vessell

3. Tube-side venting of tue isolation condenser, in plants so equippedt

4. The effect of noncondensibles in KPCI/RCIC turbine steam in plants so equipped.

These items are addressed in the following paragraphs.

As described in the Owners' Croup position, a number of provisions for venting the reactor system are available in BWR's. Ind4?d dual utilities have described their own provisions in their NUREC-0578 implementat;;n commitmen: letters.

The most ef fective of these are the saf ;y/ relief valves (S/RV). The High Pressure Coolant Injection System (RPCI) and the Reactor Core Isolation Cool-ing System (RCIC) also provide substantial venting via steam flow to their drive turbines.

Analyses of inventory-threatening events with very severe degradations of system performance have been conducted. These were submitted to the NRC 1

8-1

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I s

NEDO-24782 l t

(

• Bulletins and Orders Task Force on November 30, 1979. The fundamental conclu-sion of those studies relative to this dicussion was that if only one l

ECC system is injecting into the reactor, adequate core cooling will be provided (and the production of large quantities of hydrogen will be avoided). f t

The small-Bresa Accident (SRA) guidelines emphasise the use of HPCI/RCIC as a '

first lire of defense for inventory-thteataning events which do not quickly depressurite the reactor. If these systems succeed in maintaining inventory, it is desirable to le ve them in operation *until a decision to proceed to cold shutdown is made. Thus the reactor will be vented via HPCI/RCIC turbine steam Termination of this mode of operat-being discharged to the suppression pool.

tion (venting) would also terminate inventory makeup, and would require

~

'heactordepressurizationviathe3/RV's,whichofcourseisanothermeansof  !

venting. Termination of HPC1/RCIC is undesirable once a safe, stable condi--

tion has been achieved, until a decision is made to proceed to cold shutdown'. l

' Therefore, terminating HPCI/RCIC is not at. acceptable way to terminate venting. e

i. t If the HPCI/RCIC are unable to maintain inventory, the SBA guidelines call for use of the ADS or manual S/RV sctuation to depressurise the reactor so that ,

I The Emergency ,

the low-pressure systems (condensate, LPCI, LPCS) can inject.

ProcedureGuidelinescurrentlyinpreparationcontainthesameinstructions.

Thus, the reactor would be vented via the S/RV's to the suppression pool for such an event. Failure to depressurise,could lead to inadequate core cooling, as could repressurization above the shutoff heads of the low-pressuie systems.

Therefore, termination of this mode of venting is also an undesirable action, i and it is not recoinnended, i

If HPC1/RCIC and safety / relief valves are operable, they should therefore be e

devoted to their normal core cooling functions and be allowed to vent the reactor; it is more desirable to establish and maintain core cooling than to avoid venting. If the HPC1/RCIC and safety / relief valves are not operable (a very degraded and extremely unlikely case), another means of venting the l

reactor must be used. It is emphasised, however, that such venting is in >

the interest of establishing core cooling and therefore should not be avoided.

8-2

NED0-24732 7 ;- It it thus concluded that there is no reason to interfere with ECCS operation

[.f

'to avoid. venting. It is further concluded that the Emergency Procedure Guide--

~

- lines, by' correctly specifying operator actions for HPCI, RCIC, and'S/RV

-operat. ion, also correctly specify operator actions to vent the reactor. -

In plante equipped with isolation condensers, tube-side venting is possible to purge noncondensibles from the high point of the system. Should there be a large production of noncomiensible gases in the reactor system, it may be deemed desirable to vent the tube side of the isolation condenser to ensure its effectiveness. Per the analyses of degraded cooling referred to above'..

thiscircumstancecouldariseonlyifgocoolantinj6ctionhadoccurredforan o

extended period of time; and there would be a need to vent the isolation con-d4nser only if it were the only available means to depressurias the reactor ,

_, systetr. The need to do so is therefore quite unlikely, but if necessa'cy it coi M be hne. In plants whose tisbe-side venting is directly to the yrimary cotae!.nent, the process would be straightforward. In plants whose tabe-side

. venting .1 to the mein-condenser, such venting would be conceptually sipjlar to reopen.as a main staan isolation valve to depressurise the reactor; boo..

are possible, but should.only i used if all other approaches have f ailed.

This is because the presence of a large amount of hydrogen implies the releas-of fission products from the fuel which should not be transported-to the main-condenser.

i The effect of noncondensibles in the HPC1/RCIC turbine steam must ce con-sidered for three' cases:

1. Continuous evolution of noncondensibles due to radiolysis;

.2. Quart-continuous evolution of noncondensiblea due to core heatup;

3. LThepresenceof3quantityofnhncondensiblesinthereactorat the-time of HPC1/RCIC startup.

~

Case 1 is a normal operating mode for HPCI and RCIC and is of no concern..

8-3 l

.c-7 NEDo-24782'

_ I t'] %- e.

For Case 2, the core unst-be uncovered. However, relying again on the degraded f, I.

cooling analyses, core uncovery will be prevented-(or cladding haasup into-

~

d) if only one ECC system is operating'.'

- the rapid oxidation range will be prevente For a small' pipe break or a loss' of feedvater. which would allow the reactor .

to remain at pressure, the EPCI and/or RCIC pumps would maintain inventory and

-there would be no substantial hydrogen production. If RPCI/RCIC could not maintain inventory, the reactor would be automatically or manually depressurised -

via safety / relief valves (or via the break, for larger breaks). Depressurization; causes the HPCI and RCIC to be shut off, so the presence of noncondensibles would be no-longer an issue. '

p e performance of EPCI and RCIC under Case 3 is of ee.Lern only if there has been' a very substantial productio'n of hydro' gen due to core uncovery and th'ere

~~ *is a need to start the EPC1 or RCIC. This is an unlikely and int'olerable cir- "

=cumstance, because it could arise only if the ccre were allowed to remain uncovered for a long period with the reactor at high pressure. Automatic ,

depressurisation system operation and very explicit instructions in current "

operating procedurer, and the operator guidelines are intended to preclude thist'if the levoi has fallen with the' reactor at high pressu're, it would be depressuriud .aither automatically or manually to permit kw pressure-injection independent of HPCI/RCIC performance state. Nevertheless, the subject of HPCI operability and performance with a large volume of noncondensibles in the steam supply has been studied. The results of the study are presented in Appendix D.

8.2 REQUIREMENTS 01 PWR VENT SYSTEMS The October 30, 1979, letter of H. R. Denton (NRC) to all operating nuclear power plants contains a clarification stating that the BWR venting systems (both rearn and isolation condenser) should meet the same requirements as

. the PWR venting systems. The fo11c71ng presents generic BWR capabilities for reactor venting via safety / relief valves by comparison with the PWR require-ac.,ts in the October 30 clarification letter.* I

• Each utility must adapt this discussion plant-specifically. Isolation condensers should also be similarly addressed.

8-4 h = = ' ' ' .. .. . . .

.s .;

WED0-24782 .

(;

e The reactor vessel head can be vented near the top via the S/RV's.*

o- The size of the vents is said in the clarification to be "not a critlcal' issue." In the BWR the 'sise of: the S/RV's it, of course, determined by normal operation and safety considerations, and their capacity is substantial.' ,

e The S/RV's are not smaller than the definition of a srtli LOCA; how-ever, they discharge to the suppression pool. Inadverte.nt actuation.

is a design-basis event and a demonstrated controllable transient.

A block valve is not desirable, as discussed in Appendix-A.

+

i

,e An indication of valve position is provided in the control room.:

This indication is being upgraded'in accordance with NUREG-0578 Requirement 2.1.3A. in e Each-valve'is remotely operable from the control room, e Each valve is seismically qualified.

-e The S,/KV's are safety grade per the requirements accepted when the I

plant was licensed.

e Block valves are not required t so bloir, valve qualificationa are not' applicable.

e Inadvertent actuation is of coptce undesirable, but since the S/RV's serve an important protoctAve. function, no steps such as removal of power during poraal oper.cion<should be taken to prevent inadvertent actuation.-

l 'l e' }'he S/RV'a vent to the containment suppression pool, where discharged steam is condensed without causing a rapid containment pressure /

temperature transient.

• Big: Rock Point and Dresden I should mention steam drum.

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' s As stated above, inadvertent actuation is a design-basis event.

Temperature sensors in the S/RV discharge lines detect leakage.

3 a

No naw 10CFR50.46 conformance calcul,tions'are required, since o

4 systems in the plant's original

• design and covered by the original ~

design bases are used.

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<- S. INFORMATION REQUIRED IN TECHNICAL SUPPORT CENTER The.-owners' Group authorization for NUREG-0578 support included a study of the Because

- information to be provided in the on-site Technical Support Center.

of the evolving nature of nuclear power plant control room instrumentation and' '

' data processing needs, in response to USNRC positions which have been develop-ing,since the issuance of NUREG-0578, the Technical Support Center work has been incorporated'into a larger study of control room' functional requirements Formal reporting of the requirements awaits sponsored by the-Owners' Group.

completion of the larger study.

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, NEDO 2082-L.

2 APPENDIX A

( . .

CCNCERNS RELATED TO PRIMARY CONTAllt(ENT ISOLATION LOGIC

~1. Concern , , . .

RWCU suction and return line isolation valves are currently provided with only one containment isolation signal in addition to the process isolation signals.

Response

~WCU systes' intentionally reanins active to keep cleansing the vessel during the situation where high drywell pressure exists becaus'e the

, drywell cooleta are not operating or a small break LOCA occurs. Th6 small break LOCA could also re, in a high drywell pressure condition without reaching a low reactor ves 4. level condition. It is desirable to keep ihe RWCU operating under'these conditions.

o 2; Concern HPCI and RCIC steam exhaust line drain isolation valves are presently closed only upon trip of their, respective turbines which respond to process signals, not containment isolation signals.

Response

These systems are used to mitigate the consequences of a LOCA and thus it would-not be prudent for these valves to close upon sensing.a LOCA condition. This is also true of other valves in the HPCI and RCIC systems. The lines connecting the barometiric condenser to the Radwaste

. System could be provided with diverse and separated isolation signals, however.

3. Concern LPCI and core spray test valves are not provided with automatic isolation.

l A-1 l

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Response

Since th,e.se 1-inch valves are normally closed, fail-closed = valves, and are only opera'ed to equalise pressures to' permit testing the check valves, no

~

other isol.ttion is considered necessary. In addition, the valves are only

-opened by an operator holding a momentary pushbutton switch in'the open p- $ tion.-

o

4. Concern geactorBuilding(RS)CoolingWater/Drywell-(DW)ChilledWater.valvesare ]

not isolated.

4' e

Response

l ,

These valves are.in closed systems which are not'open to either the reactor or the crywell. In the event of an LOCA, failure.of the RB Cooling System pipirs or DW Chilled Water System piping would have to occur to allow any communication between the reactor or the dryvell and, the area outside the

~

i containment. Therefore, no automatic isolation is required'.

i 5.- concern:

<The TIP' drive line isolation valves are not provided with' direct,' auto- '

matic isolation signals.

I' Response ,

Since the TIP lines are considered instrument lines, the normal provisions

[

of 10CFR50 Criterion 56 are not applicable. The TIP's are normally with- . l drawn and the ball valves closed. S,hould an event occur while the.TIP is .

inserted-into the. core, and should the TIP fail to retract, the shear valve can be manually operated to provide the necessary containment isolatinu. /,

s i

A-2

- . , . - . - - . . - - - >f

W . . . . . _ .

4 WEDO-24782

, :1 l6. Concern g.

The CRD insert and withdraw lines are not designed to isolate after an LOCA. .

Response '

The CRD insert and withdraw lines are ,not part of the reactor coolant pres-

, sure boundary since they co not directly communicate with the reactor coolant.. The classificacita of these-lines-is quality group B and they are therefore designed in ac ordance with ASME Section III - Code Class 2.

The basis to which the CRD insart and withdraw lines are designed is commen-surate with the safety importance vf maiataint%4 pressure integrity of these

' lines .

In the design of the CRD system. lit has been accepted practice to minimise the number of valves for isolation purposes as this introduces possible .

failure mechanisms into the shutdown (scram) functien. The control roa drive insert and withdraw linas can be isolated by the solenoid valves out-side the primary containment. These lines are small; and termicate in a system that is designed to prevent leakage. Solenoid valves normally are' closed but-open on rod movement and during reactor scram. .In addition, a ball check valve located in the control rod drive flange housing automati-cally_ seals the insert line in the event of a break. Primary containment overpressurization will not result.from a line break in the primary contain-ment since these lines contain small volumes at low energy levels. -

7. Concern The butterfly valves of the reactor building to torus vacuum breaker'are not treated as safeguard equipment.

Response

These valves. perform the dual safety function of containment isolation and torus vacuum relief. They should be separated in order to maintain the intended redundant features.

l A-3 s  :

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NEDO-24782E o

'*, 8.. Concern-

_( - .

to the ,

. The isolation valves in the vacuus breaker lines which connect HPCI and RCIC turbine exhaust lines are not provided with diverse j

'i solation' signals.

Response .

These valves presently isolate only on high drywell pressur,e besides the process signals. The normal diverse isolation signal is low water level;.

however, these valves are required to be open when the EPCI and RCIC sys-taas are in u'se which would be the case for low reactor vessel water level.

- . A ' concern Isolation of the standby liquid control systea 'njection line is provided by only two simple check valves.

4 Rrsponse ,

Check valves are used for isolation in this instance so that standby

' liquid control injection will be possible even if the reactor is, isolated because of an accident. Check valves are considered an adequate isolation means and no other isolation is required besides the check valves and the normally closed explosive valves upstreas of the check valves. It is recognized that the check valve isolation scheme used on older plants does-It is, not meet the present criteria specified in 10CFR50, criterion 55. I however, felt that the inboard and outboard check valves along with the protection supplied by the normally closed er. plosive valves, provide adequate containment isolation. .

1C. Concern .

' Instrument lines connected to the primary containment are isolated only

by manual, local shutoff valves.

. A-4 m

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-'., Response-shut down 1

Automatic isolation of the instrument 111nes would, in efftet, In

.all plant instrumentation at a time when it would be most needed.

addition, t'h'ese 1ines are small which limits the communication'to the-

~

. reactor building. In order to comply with 10CFR50 crite' ion 56,' these

. lines could be fixed with one remote manually operated isolation valve per line.

r

11. Cor..:ern No automatic provision exists for-openir.3 ths RRR pump suppression pool suction valves if they are closed for shutdown cooling.

~ - # Response

'These valves are manually controlled with keylock switches. The' valves

-are normally locked open, ready for LPCI. operation, and are only closed when required for shutdown cooling. This is an intentional, deliberate action and is only performed when the reactor is shut down, depressurized and cooled down. The LPCI systam is not needed under these conditions.

Therefore. General Electric feels that no action is req 0 ired for this situation. an'd that adequate safeguards exist to ensure that the valves are in proper LPCI alignment when needed. ,

12. Concern In SIL 131 dated '3/31/75. it was recommended to delete the high.drywell-

'- pressure isolation signal for the RHR shutdown cooling suction and return.

line isolation valves. ,

Response

So as not to defeat the diversity of isolation signals which now exists, it might be advisable to retain the diverse isolation signals and utilize a keylocked " override" switch if it.is desired to place the:RHR system in shutdown cooling while a high dryvell

• pressure condition exists.

f I

A-5/A-6 i

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NEDd-24782 1 ,

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i; APPENDIX B RADIATION SOURCE TERM INPORMATION PER NUREG-0578 REQUIREMENT 2.1.6B Contents (see also Section 6)

- .i P, age B-2:

'. Fraction.~of core inventory released to reactor.

coolant (per Regulatory Guide 1.7)-

Page B-3:

,[ '

Time intervals'after shutdown (sec) at which.

activities are calculated.

4 a PagesB-4throughl8-42: Activity of each isotope in coolant (Ci and

~

grasratoms) (60 see to 8 hr).

Pages B-43 through B-45: Summary of information (60 see to 8 hr).

i ,

' [. Pages'B-46.through B-84 Activity of each isotope in coolant-(Ci and gram-atoes) (12 hr to 100 days).

Pages B;85 through B-88: Summary-of!information-(12 hr to 100 days).

D B-1

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% s j TME ISOTOPES ARE FRACit0NATED AT SHUTDOUN SY THE FOLLOWttle PERCENTS ZN 72 1.00 SA 72 1.00 OE 72 1,00 2N 7,9 1.00 OEe 78 t.00 OE 75 1.00 OA 73 1.00 OE 73 1.00 OA 74 OE 77 1.00 AS 7S 1.00 OA 76 1.00 OE 78 1.00 AS 78 8.00 SE 78 1.00 GE 74 1.00 OA 79 1.00 AS 77 1.00 SEs 77 1,00 OA 77 SE 79 t,00 BR 79 80.00 AS 80 1.00 1.00 SE 77 8.00 OE 78 1.00 AS 78 1,00 SE 70 f.00 AS 79 1.00 OEs 77 1:00 SE St 1.00 SR St 50.00 MR 81100.90 SE 82 SE 80 1.00 BRs 30 130.00 BR SO 90.00 KR 90100.00 AS 89 9.00 SEs 81 1.00 1.00 sEs 79 1.00 SR 83 SO.OO RRs 83100.00 KR SS S OO. M SE 84 1.00 SRe 82 S0.00 BR 82 50.00 RR 82100.00 SEs 83 f.00 SE #2 1.00 BR 88 80.00 RMs 8S500.00 KR 95900.0b R8 85 9.00 1.00 BRe 84 50.00ftSe RR 86100.00 BR SS 44 50.00 RR 84100.00 AS OS 1.00 SE SS 1.00 SR 87 90.00 R8 99 MR 87100.00 RB 87 1.00 BR 86 S0.00 KR 8.00 SR 89 89100.00 #9 48 9.00 5e 96 8.00 SR OS 1.00 SE 87 1.00 1.00 Yo 89 1.00 Y 89 f.00 SR SS 1.00 SR 90 1.00 y 89100.00 t.00 BR 89 90.00 KR

, ZR 90 1.00 KR 91100.00 ftB On 1.00 BR 90 50.00 Est 90100.00 RB 90 SR 92 t.00 Y 92 1.00 SR WI 1.00 Yo 91 1.00 Y 95 90 3.00 1.00 ZR 92 1.00 ZR 98 NB 93 1.00 KR 93100.00 RB 93 I.00 MR 94100.00 R8 94 1.00 SR 94 8.00 Y 94 0.00 2R 94 1.00 Y 93 1.00 ZR 93 t.00 see 93 f.00 SR 93 1.00 KR 92100.00 85 92 1.00 KR 93100.00 R8 95 1.00 SR 93 1.00 Y 95 1.00 ZR 95 ZR 96 1.00 See 94 1.00 fe 34 1.00 fte 94 t.OO t.00 te 96 1.00 MO 98 1.00 NBe M 1.00 De 95 3.00 PRO 95 f.00 Y 96 1.00 1.00 te 97 1.00 M6 97 3.00 2R 98 f.00 1.00 MRNBe 98 97100.00 fl0 97 1.00 Sp 97 1.00 Y 97 I.00 2R 97 NB 99 1.00 MO 99 1.00 MB 98 f.CO v10 96 1.00 ges 107 1 g0 1.00 TCe 99 1.00 TC 99 1.00 RU 99 1.00 TC 98 0.00 RU 98 1.00 fe 101 1.00 t9S 101 1.00 TC 101 1.00 RU IOl 1.00 MB 100 1.00 P90 800 3,00 TC 100 2R se 1.00 RU 103 1.00 RHe103 1.00 RH 103 1.00 995 102 1.00 TC 102 1.00 WU 902 1.00 000 100 1.00 NO 10S 1.00 TC 105 1.00 #95 904 1.00 TC 104 i.00 ftU 104 1.On MG 103 2.00 TC 103 1.00 RH IOS 9.00 RU SOS 3.00 RNetOS 1.00 RH 10S 1.00 RHet04 1.00 RN 804 1.00 P15 104 1.00 1.00 PS 106 1.00 TC 307 1.00 Pli 10S 1.00 TC 908 RH 108 1.00 RU 107 1.00 PO 109 1.00 AG 108 1.00 CD SOS f.00 ftM 107 1.00 PD 107 f.00 RU 106 1.00 t#tn108 1,00 f.00 AO 907 CD 109 1.00 RH 110 3.00 PD 130 1.00 AGetto 1.00 AG 110 1.00 POe109 1.00 PO 109 1.00 1.00 RH 109 TC 108 1.00 ftp 108 f.00 A0stil 1.00 AO 111 1.00 CDelli 1.00 1.00 CD 110 f.00 CD til 3.00 PD 112 1.00 AG 112 1.00 R99 181 1.00 POett3 't.00 PD 111 1,00

'Asefoe 1.00 As 100 1.00 CDett3 1.00 CD 113 f . 00 PD 884 1.00 CO 832 08AG 115 1.00 CoettS 9.00 CO IIS 1.00 1.00 AS 114 1.00 CD 114 1.00 IN tid 3.00 PO tf3 .1.00 A9e113 1.00 blNetts 1.00 sta 158 l'.00 SN 198 1.00 PO INet16 t.00 IN 118 1.00 Sto 115 1.00 SN tid 1.00 PO IIS 1.00 Ae*119 1.00 ff7 f.00 AG 117 1.00 CDe117 1.00 PD 118 1.00 AO 888 SNell7 1.00 SN 117 1.00 AG 118 1.00 CD 118 0.00 IN 918 9.00 1.00 CD 197 1.00 INett7 1.00 CD 118 1.00 SNst19 1.00 SN 119 9.00 CO 120 1.00 SN 918 1.00 CO 319 1.00 INet19 1.00 89t 117 1.00 SN 121 I.00 38 928 9.00 SN 122 IN 120 1.00 SN 120 1.00 CD t28 1.00 1N 119 1.00 1.00 8888128

$8 123 1.00 SN $24 I.00 388124 1.00 28e922 1.00 38 122 9.00 TE 122 1.00 fM 123 1.00 IN 121 1.00 900e129 f.00 TE 12S 1.00 SN 126 1.00 $88126 1.00 SS 124 1.00 TE 124 3.00 SNet2S 5.00 SN 125 1.00 SNet2S 1.00 980 929 t.00 1.00 SS 128 1.00 TE 128 9.00 SN 827 1.00 89 425 8 127 80.00 SN 128 1.00 $88128 1.00 1.00 TE 128 5.00 SB 127 1.00 Test 27 1.00 Test 29 1.00 Test 29 f.00 TE 129 f.OO 38 (26 f.00 1.00 TE v27 1.00 SN t31 1.00 $8 131 3,00 TEsist I 129 60.00 ME 129100.00 SN 130 1.00 38t 128 1,00 Tg ist 130 50.00 ME 128100.00 088 129 1.00 1.00 TE 130 1.00 SS 129 1.00 TE 132 1.00 1.00 t f.'e0 90.00 ME 130f00.00 ME 133100.00 CS1 133 132 90.00 ME 132100.00 SN 133 1.00 SS1 133 131 90.00 t.OO ME=131100.00 Teel 33 1.00 ME 131100.00 SN 832 1.00 SS 132 1.00 1.00 58 134 1.00 TE 134 1.00 TE 13't 1.00 38 835 1.00 TE 135 1.00 l 13.s 90.00 ME 134100.n0 CSof34 4.00 CSt 133 50.00 MEe133900.00 131 1.00 9A 134 1.00 ME 136100.00 CS 136 1.00 SA1 138 138 00.00 1.00 ME*13S100.00 ME 13S100.00 CS 835 1.00 Bast 3S 1.00 BA 135 f.00 ME 138100.00 CS 138 1.00 SA 138 1.00 I 137 SO.00 ME 137100.00 LS 137 1.00 BAsf37 1.00 P.A #37 1'936 60.00 CS 140 1.00 BA 140 1.00 t.A 140 1.00 CEi f39 S0.00 ME 139100.60 140 4.00 ME 949100.00 CS 141 CS 139 1.00 BA 139 1.04 8.00 1.439 f.A #39 f.00 ME 140900.00 S0.00  !

PR tot 1.00 ME 142100.00 CS 842 1.00 BA lot 1.00 LA 944 1.00 CS 143 1.00 BA 143 1.00 LA 143 1.00 BA I42 f.00 LA 142 f . 00 CE 142 1.09 PR 142 1.00 NO I42 CE 948 1.00 LA 144 1.00 CE 144 1.00 PR 544 1.00 CE 243 I.00 PR 143 1.00 NO 143 1.00 ME 143100.00 190 146 1.00 CE 147 1.00 PR 147 1.00 NO 144 1.00 CE 145 1.00 PR 145 3.03 NE 14410c.OO CS 144 f.00 SA 144 1.00 1.00 NO 145 e.00 CE I46 f.00 PMel46 1.00 Pts 148 t.00 SM 148 1.00 NO 147 1.00 PM 1.57 1.00 SH 147 1.00 CE 148 PR 946 1.00 980 181 1.00 NO 149 f.00 Pet 149 1.00 SM 149 B.00 PR 148 f.00 ND 146 1.00 1.00 PM 158 1.00 SM 151 1.00 980 ISO I,00 PPS ISO 1.00 SM 160 Set 154 1.00 EU iS4 1.00 00 IS4 1.00 EU ISI 1.00 Pet IS2 1,00 SN 152 1.00 09 153 1.00 1.00 00 157 1.00 EU IS8 1.00 00 158 1.00 SPI ISS 1.00 EU 155 1.00 00 Vs3 Et* 853 1.00 Pet 154 1.00 1.00 iEU R*t S e DY 160 1.00 00 tal 1.00 TB 161 1.00 EU 159 I.00 00 159 1.00 TB SS9 1.00 esO 158 1.00 EU 157 1.00 MO '...) 00 ISO 1.00 78 184 1.00 OY 164 I.00 DYs16S 1.00 OY 161 1.00 00 162 1.00 TB 182 1.00 OY 1sz TB 180 1.00 U 239 1.00 NP239 1.00 1.00 OY 165 1.00 H3 16S I.00 OY 166 1.00 He= W 1.00 TB iss 1 e -

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- DECAY P0lMTS - SEC0fSS 7.200E 03 1.800E 04 2.SSOE 04 3.000E O2 -6.000E 02 1.800E 03 3.600E 03 S.000E 01 1.200E 02 .

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ACTIV17Y AFTER SHUTDOWN - CURIES 1095.0 PhfD IN'1995.0 DAYS. .AT. NO.'30 2N BWR$ 238 110 1.7 PER ATL-8193 5.0 MIN. 10.0 MIN. 30.0 MIN. -80.0 MIN. 2.0 HR. 5.0 1st. s.0 1st."

ISOTOPE SHUTOOWN 80.0 SEC. 2.0 MIN.

1.048E-03 1. 047sE -03 1.047E-03 1.04SE-03 1.040E-03 1.032E-03 1.017E-03 9.727E-04 9.30EE-04 ZN 72 1.048E-03 S.419E-07 1.6SSE-11 1.54SE-20_ O. O.

ZN 73 1.774E-02 1.254E-02 8.871E-03 3.137E-03 S.S46E-04 1.041E-03 1.032E-03 1.017E-n3 - 9.7271; 9.302E-04 TSTAL 1.879E-02 1.359E-02 9.918E-03 4.183E-03 1.500E-03 CONCENTRATION AF*ER SHUTOOWN - GRAf1 ATOMS 1.300E-11 1.532E-11 1.SO9E-11 1.443E-11

'ZN 72 1.SSSE-11 1.SSSE-11 't.SS4E 1. SS".E - 1 1 ,1 '. SSI E- 11 1.S44E-11 1.643E-31 0. S.

2N 73 1.887E-13 1.334E-13 9.43SE-14 3.336E-14' S.898E-15 S.764E-18 1.761E-22 1.SO9E-11 1.443E-11 1.3eOE-11 T8TAL 1.574E-11 1.568E-11 1.564E-11 1.SS7E-11 1.SS2E-11 1.S44E-11 1.S32E-11

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ACTIVITY AFTER SMUTOOWN '- CtNttES I

.'9WRS 234 fte 1.7 PER AT1.-5193 80.0 MIN'. 2.0 let. S.O OR. S.O pet.

2.0 MIN. 5.0 MIN. 10.0 MIN. 30.0 MIN. i ISOTOPE SHUTDOWN 80.0 SEC. 7.029E-OS' 7.030E-OS. 7.030E-OS 7.030E-OS 7.032E-OS '

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6MNT1tATION AFTER SNtfTDOWN - 90 TAM ATOMS -.!

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N. 98 S.099E-09 S.898E-09 S.093E-09 S.eSAE-09 S.eS9E-09 S.811E-09 1.990E-09 1.7' HIE-09 1.SS4E-09 2.224E-09 2.199E-09 2.172E-09 2.1SSE-09 2.f47E-09 1.812E-07 2.074E-09 1.717E-07 1 *.5:ME -07, 9. 1.349E-07 N8a 97 1.831E-07 2.197E-09 1.7stE-07 fe 97 1.e40E-07 1.440E-07 1.039E-07 1.43SE-07 3.024E-10 1.CS3E-14 1.019E-20 9.357E-33 n. w ;

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NB 100 2.033E-12 1.93eE-te 2.OS3E-09 1.041E-09 S.207E-10 S.SOeE-11 1.2SSE-64 1.269E-04 1.269E-04 1.299E-04 1.20eE-04 1.20eE-04' i

te 101 1.289E-04 1. 2898E-04 TOTAL 1.269E-04 1.289E-04 i J

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1.7 PER ATl.-8193 ACTIVITY AFTER SHUTDOl#1 - CURIES 1095.01980 IN 1995.0 DAYS. 'AT. NO. 43 TO htNIS 238 ' .

9m.. 8.0 1R.

-q -

5.0 MfN. 10.0 MfM. 30.0 MIN. 80.0 MIN. .2.0 ISt. 5.0 I ISOTOPE SHUTOOWN S0.0 SEC. 2.0 MIN.

1.181E-11 1.181E-11 1.181E-11 1.181E-11 1.161E-11' S.181E-11 1.181E-11 ~1.161E-11 1.181E-11 1.181E-11 TC 98

'sI TCe 99. 4.443E O2 4.443E O2 4.443E O2- 4.44 K O2 4.443E O2- 4.443E O2 4.441E O2 4.434E O.-

4.742E-031.4.743E-03' O2- 4.3SSE O2' TC 99 4.741E-03 4.741E-03 4.741E-03 4.741E-03 4.74fE-03 4.741E-03 4.741E-03 4.742E-03 3.04SE-01' 1.974E-04. 9.54SE-10 S.324E-31 O.

O. O.

TC 100 4.087E 01 3.S20E 00 TC 101 4.747E O2 4.74SE 3.94SE O2 4.74SE 02 4.73SE O2 4.700E O2 4.452E O2'2.327E-01 O2 3.2SSE Ok f.SSSE O2 S.782E 01 1.021E 01 3.92SE O2 2.7SCE-OS O.

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TC 102 4.84SE O2 -4.20SE O2 --2.14SE O2 4.309E 0; 4.953E 00 3.414E-OS 6.200E-18 9.

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j TC 105 3.294E O2 3.112E O2 2.SSSE O2 2.239E O2 1.452E 2.970E-03 O2'S.OS7E-13 2.887E Ot 1.141E-27 9. O.

q TC 106 2.269E O2 7.371E 01 2.394E of 8.209E-01 1.OS2E-01 S. OSSE-OS 2.78SE-17 O. O. O. O.

TC 107 1.373E O2 3.269E 01 7.787E 00 S.830E-OS S.31SE-17 O. O.  : O. O. O.

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• CONCENTRATION AFTER SHUTDOtN1 - SRAM ATOMS 4.SS7E-11 4.SS7E-11 ~ 4.SS7E-11 4

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TC 104 TC iCS 1.403E-OS, 1.324E-OS 1.227E-OS .9.524E-09 9. O. O. 'l TC 105 7.43SE-10 '2.41SE-10 7.849E-11 2.891E-12 9.73SE-15 1.SSSE-24 3.739E-39 O. O. O. O.

TC 107 3.52SE-10 S.399E-11 't.OO1E-11 2.703E-13 2.072E-18 7.1304-29 O. O. O. O.

S. OSSE-20 S.598E-29 O.

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ACTIVITY AFTER SNUTCOWN - CURIES SWilS 238 110 1.7 PER ATL-5193 S.9 1W.

30.9 M191. 90.9 M191. 2.9 151. S.O Ist.

2.C MSN. S.O MSN. 10.d MSN. 'O.

ISOTOPE SNUTOUW11 80.8 SEC. O. O. O. O. 9.

O. O. O. 1.790E-21 9. 9.

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TOTAL ... . . . .. .,e;- 04 9.235E-te 9.2SSE*e4 9.299E-ee 0.999E-ee 9.Ostt-ce 0 99et.es -- O.394.m o gnag .

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'94N06 23. II. 1.7 PER ATL-Sits ACTIVITY AFTER SNUTDOWN - CtNtfES 1995.9 MMS fet 1995.9 SAYS. l s 3 2.0 M190. S.O MIN. 10.9 MON. 30.0 MIN. 9 0. 9 81998. 2.. Det. S.e Det. e.9 set.

130 TOPE SHUTDOWN 00.0 SEC. 9. S. S.

SS 121 O. O. O. 9. O. 9. O.

e.

l see122 ~.9SOE-04 7 S.7deE-04 S.721E-04 3.4e7E-94 1.SteE-04 3.SteE-OS s.977E- S 1.987E-12 2.47.E-25 i

4. 930E-02 d.989E-02 4.993E-et 4.edlE-92 4.SeeE-02 4.94K-et j sa 122 4.945E-02 4.944E-02 4.943E-02 4.94tE-02 I

se 123 O. O. O. O. S. J. O. S. e. S. ,

ase124 2.esOE 00 1.701E 00 1.Oe7E 00 2.841E-91 3.034E-SE 3.949E-OS 5.SSSE-12 1.29K-23 e. e. .

- SS 124 2.95eE 00 2.95eE 00 2.958E 00 2.95eE 00 2.957E OO -2.957E 00 2.95GE e5 2.955E OS 2.981E SS, 2.94K es d SS 12S 2.527E 00 2.327E 00 2.527E 00 2.527E OC 2.S27E 00 2.527E 00 2.S2eE es 2.S2eE Ge 2.82eE es. 2.SteE 1.472E-e4 es S.2edE-Of S.059E-01 S.e42E-Ol R.23eE-Of 4.363E-01 2.104E-Of 7.049E-et S.Ot3E-03 f

SSet28 ..Sc3E-94

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• 4.4SK 01 4.437E 01 4.382E OS 4.214E 01 3.StSE Of 2.SS2E el 8.309E Of 9.75tE se 2.34tE-et Seet2e 4.47sE Of 7.997E-St SS 120 1.299E 00 1.299E 00 1.299E 00 ?.299E 00 1.29eE S.140E 00 01 1.291E 7.7e3E 80 1.272E Of 7.191E 90 el 1.21M e.121E 90 et 9.9ESE-Of 3.779E 91 2.32eE Of se 129 3.255E Of 3.149E 09 e.242E OS 0.214E Ol SS 130 1.929E Of 1.900E O2 1.SSM Of 1.eO2E Of 1.e53E Of f.139E O2 S.49eE et 2.1fet et 7.23eE-et 2.de3E-et 58 131 2.459E O2 2.40eE O2 2.352E O2 2.175E O2 1.89eE O2 1.Oe9E O2 4.74tE et e.'*SIE se S.197E-et 4.152E-04 2.875E DE 2.194E O2 1.7 2E O2 9.290E OS 3.053E Of 3.492E-Of 4.2SeE-94 S.379E-19 2.12SE-27 9.

Se 132 SS 133 1.905E Of 1.405E Of 1.16M O2 S.389E OS 1.de9E 91 . 775E-Of 3.987E-ee S.110E-12 3.9 m -St G.

s.eDeE 01 1.852E 00 3.930E-SE 4.19tE-07 2.SSOE-19 9. e. c. S. e.

38 134 36 135 3.SGSE 09 1.131E-Os 3.491E-18 O. - O. e. O. 8. S. S. .

TOTAL 1.171E 03 9.992E O2 S.e44E OS 7. OSSE St S.SSIE OR 3.752E O2 2.499E of 1.429E et 7.eeeE Of G.94K St CONCEftT9tATION AFTER SMUTDelet - enen r* cme .

i SS 121 1.290E-OS 3.290E-05 1.2SOE-05 1.2SOE-05 1.2SOE-05 1.281E-05 1.281E-OS 1.29tE-es 1.Selt-ce 1.2c1E-es See122 1.777M-14 1.SO7E-14 1.27eE-14 7.700E-15 3.49tE-15 1.25eE-96 4.379E-19 4.43eE-23 S.53M-Se 1.OS2E-09 1.OO'E-09 9.99fE-te G.

S. S.K-19 SS 122 1.022E-Os 1.022E-OS 1.022E-OS 1.et2E-09 '1.OtlE-89 1.017E-99 t SS 123 2.77X-OS 2.773E-OS 2.773E-05 2.773E-05 2.773E-05 2.77M-95 2.~773E-95 2.773E-05 2.773E-e5 2.773E-85 y

1.40tE-tI S.95eE-12 2.34tE-12 2.SOOE-13 3.253E-17 4.930E-23 5.OS4E-34 S.

-e-s 38e124 2.192E-fl 8.353E-OS 3. 3SM-Os 1.383E-Os 1.383Z-OS 1.383E-OS 1.363E-OS 3.363E-ee 1.382E-ee 1.SeeE-ee 1.3eeE-ee 5

4 U SS 124 1.907E-OS 1.907E-05 1.907E-05 1.907E-O"J 1.907E-OS 1.907E-95 1.907E-OS 1.900E-95 C 125 Sest28 1.907E-05 S.347E-31 1.907E-05 S.s20E-1, 8.900E-11 S.299E-11 4.407E-11 2.12SE-is 7.s20E-12 3.se3E-33 s.59eE-s4 1.4eM-i4 3

e. 21M-ee 0.19eE-te S.142E-es e. ecee-ee Se 32s 3.230E-Os e.230E-te e.230E-Os e.229E-OS e.229E-ee S.225E-Os 9. 24M-97 9. 2t M-e7 9.90eE-97 e.seeE-97 38 127 9.2GSE-07 9.2GGE-07 9.2eSE-07 9.2SSE-07 9.264E-07 9.259E-07 1.eS9E-OS 1.357E-te S.982E-19 9.38SE-11 1.2 1E-11 i Sdette 2.342E-09 2.378E-09 2.359E-OD 2.399E-09 2.241E-09 38 128 3.730E-09 3.730E-OS 3.730E-09 3.729E-09 3.727E-OS 3.709E-09 3.SSM-ee 3.deOE-99 2.350E-99 2.271E-ee 38 129 1.133E-07 1.132E-07 1.13tE-07 1.12M-97 1.110E-07 1. OSSE-e71.2 9.SeeE-ee 8.deeE-ee S.leeE-te 3.19M-ee 4.es -it 3.795E-0. 3.75M-Oe 3.7 m -0. 3.544E-Oe 3.2SiE-ee 2.24x-Os e.3 m -ee 1.i.4E-.9 .i ..42M-1.

m -se 4.i m -99 SS i30 i.44eE-u. l i M-i 2 S.SiK i4 3 13 3.=9E-OS 3.202E-OS 3.i m -OS 2. m -se 2.52:E-.e SS 132 4.4 TOE-09 3.S17E-09 2.937E-09 1.53tE-99 S.033E-19 S.757E-12 7.03eE-15 1.05tE-te3.194E-22 3.Se3E-Se c.

S.

3.

SS 133 2.735E-09 2.131E-09 1.seeE-OS 7.700E-19 2.13eE-tO 1.2SOE-12 S.eeeE-te

e. e.

S.637E-11 1.513E-12 3.449E-14 4. OSSE-19 2.814E-27 e. c. S.

SS 134 S.87M-92 3.904E-21 S.37SE-31 e. 9. e. O. S .' O. c.

38 135 1 TOTAL S.190E-05 e.19et-05 S.197E-05 S.197E-05 e.19eE-OS S.193E-05 S.19eE-SS S.197E-05 e.leEE-SS S.17eE-95

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e ACTIVITY MTER SHUTOSINS - CURSES 1995.0 lede 1N 1995.9 env3. AT. Ose, el Pp1 S4416 238 HS 1.7 PEll ATL-9193 4 -

ISOTOPE SHUTOObsst 90.O SEC. 2.O MSN. S.O MIN. 1O.O M111. 30.e 79191. 90.e 991$1.' 2.O 981 S.e 981. e.e 981. r Pr1 147 S.SeeE 01 S.SOSE 01 S.SOSE 01 e.SeSE 01 S.SeeE 01 S.SeeE 01 'S.SeeE 01 c.907E 01 S.997E el e. ESSE el pytelde 4.eO7E 00 4.OO7E 00 4.eO7E 00 4.eO7E 00 4.eO7E CO 4.eOSE 00 4.004E'OO- 4.eOVE OO 4.7 1E se 4.761E et PM 14e 3.830E 01 3.630E 01 3.529E 01 3.829E 01 3.627E 01 3.820E.01 3.811E 01 3.592E Of 3.535E el 3.479E el -

PM 149 1.'474E O2 1.374E O2 1.374E O2 1.37M O2 1.373E O2 1.371E O2 1.3SSE Of 1.357E et 1.31eE St 1.200E et Pet ISO. 1.72SE-01 1.71eE-01 1.710E-01 1.SeeE-01 1.SS3E-01 1.517E-01 1.334E-01 1.032E-01 4.779E-et R.212E-et

• PM 151 S.SSSE 01 S.SS7E of S.SS7E 01 S.SSSE 01 S.SSIE n1 S 619E Of S.SSSE el S.421E el S.e3M el 4.673E et Pf1 182 3.959E 01 3.527E 01 3.142E 01 2.221E 01 1.247E Of 1.23eE pp 3.eS2E-Of 3.767E-95 3.49eE-14 3.247E-23 PM 194 1.413E 01 1.071E 01 0.11M 00 3.SSSE 80 S.SteE-01 3.447E-03 m.413E-97 S.010E-14 e. G.

T37a 3.SSeE O2 3.de1E O2 3.41eE O2 3.27eE O2 3.153E et 3.02eE Or m.OOSE Of 2.97eE et 2.eteE 08 2.eest et cepeCENT11ATI081 MTER SMU) D1hM - SINGS ATOMS P91 147 4.902E-94 4.902E-04 4.902E-04 4.902E-04 4.9021-04 4.902E-04 4.902E-94 4.902E-94 4.eeSE-ed 4.eeK-ed i Pfle143 1.54SE-OS 1.54SE-OS 1.54SE-OS 1.SeeE-OS 1.5468.-08 1.94SE-OS 1.54SE-es 1.544E-ee 1.541E-ee 1.53eE-es l Pr1 144 1. Sole-OS 1.SO1E-OS 1.SO1E-OS 1.SeOE-OS 1.500 & O6 T.497E-OS 1.493E-98 1.deSE-86 1.de2E-98 1.43SE-98 2.324E-e8 2. 32M-98 2. 3fT.-98 2.320E-OS 2.313E-98 2.200E-OS 2.22SE-98 R.14SE-98  :

Pro 149 2.324E-OS 3.324E-OS I PM 150 1.40SE-10 1.deOE-10 1.473E-10 1.454E'-le 1.42M-VO 1.307E-10 1.1SOE-te e.es2E*11 4.117E-11 1.SeeE-11 Pf1 191 S. OSSE-07 S.05eE-07 S. OSSE-07 S.OSM-97 S.OSOE-97 S.021E-07 4.96SE-97 4. ease-97 e.-M -97 4.17eE-e7 i Pf1 152 1.283E-09 1.12SE-09 1.OO2E-OS 7.907E-10 3.977E-10 3. e4M- 11 1.232E-12 S.202E-19 1.11SE-94 1.030E-33 Pf1 154 - 1.07eE-10 1.423E-10 1.079E-10 4.SO4E-11 1.17M-11 4. sex-14 1.IteE-17 e.GSOE-RS 9. S. l T:Ta 4.960E-04 4.960E-04 4.960E 4.9eOE-04 4.9eOE-04 4.9eOE-04 4.96x-ed 4.9eOE-ee 4.es E-ed 4. esse-ed l l

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I 1995.0 MMS SW 9995.9 OAYS. AT. Ges. 93 gg gemte 236 see 1.** PER ATL-5193 ACTIVITY AFTER SMUTDOMet - CURtES 30.9 MtN. 90.9 ftlN. 2.9 Det. S.S DR. e.e OR. "

5.9 MIN. 10.o MIN. e.

2.0 MIN. e.

ISOTOPE SMUTOOWN 90.9 SEC. e. 9.

O. 9.

EU 151 O. O. O. O.

9. O. O. 9. e. e.

O. O. O. 6. S.908E-e1 S.SeeE-01 EU 153 8.987E-Ol 8.90eE-el S 99eE*e1 .S 90eE-01 S.9eeE-el S.SeeE-el EU 184 S.097E-01 S.987E-OS 1.895E 01 1.895E et^ 1.699E Of 1.895E et 1.695E el 1. OSSE 01 9.se8E et 1.ee8E et .

I EU 158 1.695E 01 1.495E Of EU 190 S.714E 00 S.794E 00 S.714E 00 S.7f3E 00 3.782E 00 S.70eE 00 9.703E OS 5.Se2E 90 9.eeeE Ge e.SeeE se EU 157 4.629E 00 4.S2SE 00 4.822E 00 4.StIE OO 4.S93E 00 4.524E 80 4.422E 004.00eE-et 4.22SE se 2.eeeE-et 3.eesE se1.798E-93 3.214E se 2.44SE 90 2.409E 00 2.373E 00 2.20eE 00 2.103E 00 1.95SE 00 9.90tE-et 9.84M-92 S.78et-93 9.53eE*ee 9.399E-99

• EU 19e 9.921E-Ol 9.94GE-01 9. lese-Of e.le3E-Of S.750E-Of 3.12SE-OS 9.

EU 199 2.473E-OS l.874E-61 0.157E-02 2.039E-U2 7.963E-OS 1.943E-OS 1.157E-15 9. 't.7eeE et f.e4K et 3.114E Of 3.07SE 01 2.97SE nt 2.SeeE 01 2.79eE #1 EU 100 3.2S X-Ol 737AL. 3.87eE 01 3.leOE Of 3.14eE 01 CONCENT1tATION AFTER SMUTOOMet - SIWWE ATOMS EU 155 S.843E-OS S.843E-OS 6.843E-09 S.643E-09 S.844E-09 1.430E-04 S.647E-09 1.SSOE-04 S.95tE-te 1.030E-94 S.eesE-te 1.e30E-04 S.SesE-te 1.83tE-04 8.70eE-ee-1.829E-04 1.429E-04 1.029E-04 1.e30E-94 f.asoE-04 EU 153 I EU 184 3.129E-OS 3.129E-05 3.129E-95 3.929E-05 3.129E-OS 3.129E-OS 3.129E-05 3.129E-08 3.lteE-05 3.129E-e8 EU 1s5 s.s75E-Os e.s75E-Os e.s79E-Os e.srSE-Os e.37SE-Os e.s74E-Os S.s74E-es e.s74E-es e.s73E-es e.s72E-es EU 198 S.SS3E-07 2.24SE-OS 2.243E-OS S.SSOE-072.241E-OS S.SSOE-97 2.23eE-Se e.649E-07 S.84SE-97 2.22eE-OS 2.994E-OS S.644E-07 2.144d-ee S.63eE-e7 2.049E-ee S.82SE-07 1.707E-te S.90eE-97 1.geeE-es s.99et-e7 EU 157 s.90eE-12 4.317E-13 EU 15e 5.99tE-10 9.892E-te 5.eOSE-te S.847E-te 9.944E-te 3.000E-te 2.42tE-te 9.eOSE-19 7.e33E-18 S.4SSE-19 2.99eE-sl 9.42tE-12 9.347E-13 9.129E-16 S.91SE-19 EU 199 9.49eE-11 9.137E-If S.792E-19 1. 53E-15 2.SSSE-19 1.530E-te 9. S.

l j EU 100 4.337E-12 3.2e7E-12 2.49tE-12 1.904E-12 2.711E-13 3.OO7E-04 3.OO7E-94 3.Se7E-04 3.007E-04 3.Seet-c4 TOTAL 3.OO7E-04 3.OO7E-04 3.OO7E-04 3.OO7E-04 3.OOFE-94 .

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90 ISS 0.

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S. e. S. se e. 8.SteE e.

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60 194 1.4SeE-ee t.49eE-ee 2.327E-93 1.322E es 1.27tEt4.aeeE-es 90 1.832E 2.829E-te se 1..eet 1.GeeE-te eO 155 tepectesTWAY8est AFTER SMSTecust - est4M 1.4SSE-te ATOMS ce 90 15e 4.4eet-97 4.494E-or 4.4esE-er 1.4 eE*ce 1.4SSE-ce -

90 197 S.Se2E-SS 8.SS2E-95 S.042E-OS 4.4e7E-87 . 4.4est-or eO 1se z.oSee-es m.e34E-e5 S.e#EE-OS 5.eO2E-OS 4.SO2E-e7 1.4 eE-te

5. OSM-99 S.888E-SS 1.4eSE-OS 4.SetE-e7 4.sseE-e71.4eet-es t.498E-ee t.det

4. arse:-or 4.7e3E-e7 i

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'30 tot 3.eeEE-13 3 2. 34E-95 2.834E-05 f.034E-ee 2.834E-95 R.S 2.834E-ee 80 in Twm. 2.137E-te 2.137E-te 2.137E-OS 2.137E-Os 2.137E-ee 3.137E 7.OneE-in o.w4E-OS .e.301E eet-Sm 2.737E-t2 12 0.143E-in

.ee4E-es e.wM-oS o.eees:-es 3.seeE-82 0.194E-13 1.442E-14 S.OneE-ie 3.eeds:-se 7.7 3E-te 7.97eE-ee 7.feeE-99 e;ee3E-ee. e.7.S4E-es S.229E-97 e.edeE-22 E.137E-et-9.e43E-Se S.

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ISOTOPf! - SNUTOOMet 90.0 SEC. 2.0 MIN. 8.0 MIN. 10.0 MSN. 30.9 hfN. 98.9 MON. 2.9 let. S.S 984. C.9 93.

TS ISW O. O. O. O. S. S. O. S. S. G.

TS 100 1.237E-Of 1.237E-01 1.237E-Of 1.237E-Of .I.237E-91 1.237E-Of 1.237E-91 1.23eE-et 1.230E-99 1.29M-el To tot 2. ecee-Ot 2.eOet-Ot 2. ecee-01 2.eOSE-Of 2.eO4E-OS 2.SetE-et 2.SeeE-et 2.9eSE-99 2.993E-St 2.St1E-Of

• T3 le2 9.4SOE-Ot" 1.44sE-OS 9.44eE-Of 1.449E-at 1.423E-01 1.397E-Of 3.13eE-St e.3etE-et 3.22eE-et 9.233E-et TS 183 S.04 M-02 S.032E-02 S.02tE-02 S.900E-02 S.93eE-02. S.729E-OE J.43tE-92 4.942E-92 3.545E-et 2.474E-et TS 164 7.eS2E-03 7.84eE-03 7.844E-03 7.83fE-03 7.et M-03 7.734E-03 7.St9E-93 7.393E-92 S.754E-93 0.179E-93 TOTA 4. S.97eE-01 S.973E-Of S.971E-Of S.9SOE-Of S.937E-91 S.GOSE-et S.Se7E-99 8.293E-Of '4.53fE-01 4.19eE-91 COpeCEpsTftATIOpe AFTEst SNUT90M80

• Sftfet ATONS TS 19e 4. sex-OS 4. sex-OS, 4.9G M-OS 4. sex-OS 4.Se X-Os 4.Se3E-OS 4.844E-98 4.984E-98 4.805E-98 4.900E-98 To teO S.e20E-Os S.020E-OS S.820E-Oe S.e20E-OS s.e20E-OS S.eleE-Os S.st7E-ee S.SISE-OS 8. ecee-ee S.79eE-ee Ta set 1.377E-Os 1.377E-OS 1.377E-Os a.37eE-Os 3.37eE-Os 3.374E'Os 1.37tt-ce 1.3eeE-Os 3.349E-te 1.332E-te TS 162 1.Oler-lO 1.017E-10 1.018E-te 1.OllE-10 9.99tE-91 9.244E-fl 7.970E-91 S.82SE-II 2.P#4E-11 9.79eE-12 TS 163 1.2S3E-SO 1.25tE-10 1.249E-19 1.242E-10 1.239E-lO 1.1seE-10 1.12eE-99 1.Ol2E-19 7.3SSE-11 S.33eE-11 TS led 5.762E-19 S.7SeE-fl S.7SeE-11 S.747E .lt S.733E-11 S.978E-St 5.59tE-ft S.425E-St 4. tr./4:- 11 4.327E-91 -

T37A4. 4.76eE-OS 4.766E-Os 4.7eeE-OS 4.76eE-OS 4.76eE-OS 4.76GE-OS 4.766E-OS 4.796E-OS 4. W.8~-OS 4.787E-est i .

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stats 23e fee 1.7 PER ATL-9193 ACTIVITY AFTEst SMUT 90Wes - cafettES 1995.0 seID th 1995.9 BAYS.

ISSTOPE SMUTOOWet 90.0 SEC. AY. See. es WY j y- DY t90 O. 2.0 MIM. S.O MIN. 10.0 MSN.

9. O. O. 30.8 Mtfe. 8 0. 0 99990. 2.9 Det.

) '

JY ISI O. S. O. - 9.

9. 9. O. *9.

S.9 set. 4.e 93.

JY fSt O. 9. 9 S. S.

9. O. O. O. S.

t DY f S3 0. 9. O. O. O.

S. S.

OY 184 O. 9. O. 9. 9.

- O. 9. 9 9 O. 9.

DYetSS O. O. O.

S. S. 9.

D7 185 3.OOSE-Of 1.73tE-Of 9.972E-93 1.90SE-03 f.207E-04 O. 9. 9.

3.947E-99 1.28tE-96 S.200E-31 S.

DY ISS 1.970E-03 1.970E-03 1.SSSE-03 9.SS9E-03 9. S.

TOTAL 3.de8E-02 3.400E-Of 3.467E-02 5.SS7E-033.499E-OE 9.SSSE-93 1.SS4E-03 3.482E-SE 1.939E-93 3.467E-02 1.7seE-93 1.747E-93 3.44eE-SE 3.

DY 100 S.SSIE-02 S.40SE-Of 2.703E-07 2.703E-07 4.671E-Ot 3.SS3E-OE 3.SetE-02 3.954E-02 3.63tE-Of 3.

Cef0CENTftAYlege AFTEft SMW7DOMIS - Oft 4M A70Pt3 CY 181 2.703E-07 DY 1St 4. 67t E-07 4.47tE-07 4.97tE-07 4.871E-07 2.703E-07 2.703E-07 2.794E-07 2.794E-97 2.794E 97 2.798E-97 f.700E-97 I

DY los 8.22SE-07 3.303E-07 8.229C-07

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HR. 1.0 DAYS 3.0 DAYS S.O DAYS 10.0 DAYS 30.0 DAYS 90.0 OAYS 100.9 BAYS 384.6 BAYS ISOTOPE SHUTDOW94 12.0 6.70SE 09 S.SeSE Of S.402E OS 5.teeE et 6.686E 01 S.690E Of S.693E of S.70SE Of S.794E Of S.72eE Of 9.2F9E-01 1.17M-03 PM 147 2.930E.00 1.786E 00 I Pfta144 4.807E 00 4.768E 00 4.728E 00 4.57SE 00 4.426E 00 4.076E 00 9.G66E-Of 1.392E-OI S.364E-Of 9.OS7E-04 3.630E Ol' 3.40SE 01 3.196E OS 2.479E Ot 1.924E 09 1.02SE OS 3.32r4*-12 e. Pft 146 1.374E O2 f.20SE O2 1.030E O2 S.499E OS 2.934E 01- S.114E 00 9 . t ** t E-02 9.39tE-07 O. S. Pft 149 O. 1.72SE-01 7.923E-03 3.639E-04 1.620E-09 7.200E-15 3.01M-Fe O. 1.900E-14 9.134E-25 S. i Pfl 150 3.14SE 01 '9.Se9E 00 2.923E 00 9. 1.500E-OI t.040E-OS PM tSt S.SSOE 01 4.233E 01 O. 'O. .O. 9. e. . 3.959E 01 O. O. O. e. S. PM 152 O. O. O. O. Pt1 154 1.413E 01 O. O. O TOTAL 3.SS8E 02 2.68SE O2 2.3SIE O2 1.610E O2 1.231E O2 s.787E 01 7.09eE 01 a.77eE 01 S.Sett et 5.teet el CONCEttTRATIONAFTERSHUTOOWet-GRhMATOMS. 4.9 eE-m 4. eReE-= 4. 93E-m . . .= -m 4.9:SE-= 4. 92x -= 4.93x-= PM 247 4..On-= 4.904E-= 4.907E-= 1.546E-OS 1.533E-OS 1.52tE-OS 1.47tE-OS 1.424E-OS 1.311E-OS 9.423E-07 S.743E-97 2.900E-97 3.70eE-ee Petalde 1.40eE-OS 1.322E-OS 1.02SE-OS 7.95SE-07 4.237E-07 3.997E-Os S.757E-09 2.SSIE-99 3.33eE-11 1.501E-OS 4.988E-07 1.034E-07 8.94eE-10 1.SeeE 5.824E-RO e.- PM 14e 9.3 ole *07 Pts les 2.324E-OS 2.03eE-OS 1.742E-OS O. . S. 9. PM 150 3.13SE-93 1.39SE-IS 6.209E-24 2.895E-37 6. e.  ! 1.406E-!O S. OSSE-07S.82SE-12 3.782E-07 ' t.ellE-07 S.SSOE-OS 2.692E-08 O. 8.340E-09 9.2952-15 1.997E-22 0.192E-39 S. Ptt 1st O. O. S. S. PM 152 1.263E-09 O. O. O.

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O. O. O. O. PM 154 1.878E-10 4.960E-04 O. 4.958E-04 4.999E-04 4.990E-04 4.950E-04 4.951E-04 4.92SE-94 4.834E-94' 4. ecee-94 3.87K-04 TOTAL . l

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                                                                ' CONCE8978tATION 2.173E-04 AFTER SHUTOOWN      - 98 tart 2.199E-04         ATOMS 2.262E-04      2.380E-04      2.900E-04     3.394E-04 1.790E                                                                                                                f.786E-04    1.790E-04      1.793E-04 SM 147               2.1SSE*04 2.197E-04 2.189E-04               2.166E-04 1.771E-04   1.774E-O<  1.77eE-04 1.76SE-04 1.766E-04 1.767E-04 SM 148 SM 149               S.117E-OS                     8.462E-Cd 5.7S8E-08 S.7S9E-04 5.7S9E-04 S.759E-04 S.799E-04 S.759E-04 8.570E-OS 7.603E-OS 7.39eE-OS 7.499E-OS 7.

S.759E-04 S.759E-04 S.759E-94 1.OOSE-OS 1.OO9E-OS*.1.OOeE-OS S.799E-04 9.759E-04 t.OOSE-OS 1.OO7E-OS 1.OO9E-OS SM 100 1.OOIE-OS SPt 181 9.582E-06 9.713E-08 9.OIOE-OS 4.129E-04 4.129E-04 3."341E-OS 4.129E-04 2.08EE-11 4.129E-047.279E-Is 4.129E-04 4.12SE-049.4.129E-04 5.157E-22 ' SM 152 4.129E-04 4.129E-04 4.129E-04 2.079E-07 1.22tE-06 I.023E-06^ S.SS7E-07 4.220E-07

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SM 1S3 Srl 154 7.136E-OS 7.136E-OS 7.I36E-OS 7.136E-OS 7.136E-OS 7.I36E-OS 7.136E-OS 7. O. G. O. 1.419E-09 8.839E-IW 2.572E-28 f.47tE-03 1.472E-03 f.479E-03 1.da3E-03 O. 9. 1.404E-03, O. . 1.Seet-OS 1.9eeE-OS Sf* tSS 1.469E-03 TOTAL 1.468E-03 1.469E-03 ,. .* . O s, o e O w

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l ( APPENDIX C BWR OWNERS' GROUP POSITION ON HICH-POINT VENTS NUREG-0578 Implementation Letter Requirement Relative to Rosately ryerated , Operated Righ-Point Vents 1 Each applicant and licensee shall instali rasetor coolant system and reactor vessel head high-point vents remotely operated from the control room. Since , these vents form a part cf the reactor coolant pressure boundary, the design of the vents shall conform to the requirements of Appendix A to 10CFR50 General Design Criteria. In particular, these vents shall be safety grade,- , and shall satisfy the single-failure criterion and the requirements of  ! IEEE-279 in order to ensure a low probability of inadvertent actuation. , b Each applicant and licensee shall provide the' following information concern-ing the design and operation of these high-point vents:

1. A description of the construction, location, size, and power supply for the vents along with results-of analyses of loss- l of-coolant accidents initiated by a break in the vent pipe.

The results of the analyses should be demonstrated to be acceptable in accordance with the acceptance criteria of 10CFR50.46.

2. Analyses demonstrating that the direct venting of noncondensible gases with perhaps high hydrogen concentrations does not result in violation of combustible gas concentration lieits in the con-i tainment as described in 10CFR50.44. Regulatory Guide 1.7 (Rev. 1), and Review Plan Section 6.2.5.

1 l; L 3. Procedural guidelines for the operators' use of the vents. The r information available to the operator for initiating or termina-ting vent usage shall be discussed. C-1

1' d.c ,, . i f c - WED0-2482 V T '. _ 4.. Diocussion t Domestic BWR's are provided with a number.of power-operated safety-grade relief valves which can be manually operated from the control room to vent the reactor pressure vessel. The point of connection of the vent lities } from the vessel to these valves is su'ch that accumulation of gases above that point in the vessel will not affect natural accumulation of gascs of the L reactor core. ,, , , These power-operated relief valves satisfy the intent of the NRC position. Information regarding the design, qualification, power source, etc., of these valves has been provided in the individual plant Safety Analysis Reports. The Owners' position is that the requirement of single-failure' criteria for prevention of inadvertent actuat. ion of these valves., and the requirement

                                                                                                            .1 (stated in the October 11 topical meeting) that power.be removed during               l l

normal operation, are not applicable to BWR's. These valves serve an ' i important function fn sitigating the effects of transients and in many plants provide ASME code overpressure protection. Therefore, the addition of a L second " block" valve to the vent lines could result in a less safe design and ! in some cases a violation of the code. Also, inadvertent opening of a relief l valve in a BWR is a design basis event and is a controllable transient (this i ( is discussed under Item 2.1.2). l L i l In addition to the power-operated relief valves, operating BWR's include various other means of high-point venting. Information on which plants are equipped with which features has been provided-in individual plant Safety Analysis Reports, and may be sununarized by individual licensees in their NUREG-0578 implementation letters. Among these are: 1 l

1. Normally closed reactor vessel head vent valves, operable from L

i the control room, which discharge to the dryvell; l ,

2. Normally open reactor head vent line, which dischatges to a main steam line;

l. .

                   ,'                                           C-2                                            l

                                             ~'

NEDD-24782 i u.

             /L '.

3,

              ,                               Main' steam-driven Reactor Core Isolation Cooling (RCIC) System'       ,
                                             ~ turbines, operable from the control room, which exhaust to the.

suppression pool;

4. Main stese-driven Righ Pressure Coolant Injection (HPCI) System .

turbines, operable from the control room, which exhaust to the suppression tool; -

5. Isolation condenser primary side vent valves, operable from the control room, which discharge to the containment or a main steam line.

l l'  : l Although the power-operated relief valves fully satisfy the intent of the; i requirement, these other means also provide protection against the accumu-lation of noncondensibles in the reactor pressure vessel. In the October 11. 1979, topical meering on this subject,"three procedural questions were raised: 1 l

1. I Where to vent to (suppression pool vs. containment);

I

                                                                                                                   .)

2. When to vent;

3. When not to vent.

l Under most circumstances, there would be no choice as to where to vent to or l

                                                                                                                    .]

when to vent, since the relief valves (as part of the Automatic Depressuriza- { tion System),-HPCI,fand RCIC will function automatically in their designed.

                                                                                                                   .{

modes to ensure adequate core cooling, and these will provide continuous ' venting to the suppression pool. The current assessment is that it would a not be desirable to interfere with emergency core cooling functions in order to prevent venting, but the matter will be studied further. 1 1 l The result of a break in the safety / relief valve discharge line, or any of the other systems enumerated above, would be the same as a small steam line i C-3 l L ,,. .

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NED0-24782 .

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break. A co.41ste steam line break is part of the' plants' design basis, and ' smaller-size breaks have been shown to be of lesser severity. A number of reactor system blowdcvns due to stuck-open relief valves (siso equivalent to a small steam line break) have confirmed this in practice (see owners' Group position on Requirement 2.1.2). Thus no new analyses to show conformance with 10CFR$0.46 are required. Because the relief valves, EPCI, and RCIC will vent the reactor continuously,. and because containment. hydrogen calculations in normal safety analysis cal-culations assume continuous venting, no special analyses are required to l-demonstrate "that the direct venting of noncondinsible gases with perhaps I high hydrogen concentratiors does not result in violation of combustible gas concentration limits in containment." BWR Owners' Group Implementatian Critaria l 1

1. The owners' Group believes that adequate reactor coolent system -l venting is provided by the existing plant design.

2. Plant procedures will be provided to govern the operator's use of the relief valves for venting the reactor pressure vessel.

H 3 No new 10CFR50.46-conformance calculations or containment com-bustible gas concentration _ calculations are required, since sys-tems in the plant's original design and covered by the original design bases are used; l

4. In response to a request from the October 11.-1979, topical meeting, the use of isolation' condenser tube-side vents will be. -i considered;

5. In response to a-request. from the October 11, 1979, topical meeting, the effect of noncondensibles in HPCI/RCIC turbine steam will be addressed.

1; ! 1 L C-4 i

n , L 1 WED0-24782 , APPENDIX D~

      /.                .

RPCI TURBINE PERFORMANCE WITH HYDgDGEN IN SUPPLY STEAM This appendix presents the results of an investigation of the affect on arCI. l turbine thermodynamic performance of various volume concentrations of hydrogen. , gas mixed with dry steam. It is concluded that there is no degradation in , thermodynamic performance of the turbine with a mixture of steam and hydrogen. , The available energy of hydrogen is significantly greater than that of saturated q mteam at equal pressure (i.e.,12.5 at 1000 pois for 100% steam /0% H2 : 30% T steam /70% H 2 ). The specific volume of hydrogen is also significantly greater

                  -than steam at equal. pressure (i.e., 1:3 at 1000 psig for 100% steam /0% H2 '

30% steam /70% H2 ). Consequently, for a fixed mass flow the volumetric flow would increase with increasing concentrations of hydrogen. The net result of ] this ef fect is increased pressure drop in the turbine inlet and exhaust steam lines due to higher flow velocities. The following tabulation susmacizes this

                                                                ~

point: , 100% Steam /0% H2 30% steam /70% H2 1 1000 pois. 4000 bhp 1000 psis, 4000 bhp Inlet velocity 43 ft/see 150 ft/sec

                                                                                                            ~

(10 In. Inlet) Exhaust Velocity 205 ft/sec 535 ft/sec ): (18=In. Exhaust) 1

                                                    .                                                          1 Figure D-1 shows the minimum performance capability'of a typical HPCI turbine.            '

The 'Hixture Flow' line represents a 100% steam /0%-hydrogen mixture. This l line shifts upwards (rotates counterclockwise) with increasing hydrogen concentration. The higher steam line velocities identified.,might be a problem under long-term [ l' operation. However, since hydrogen gas wodid exist in the steam for only a

                     'short time, the high velocities are not considered to be detrimental.

i' L The ability of the turbines to start with hydrogen in the supply steam has als'o bee considered. With the increase in available energy due to hydrogen i

          -           gas concentration (on the order of 2.5 to 1 with 70% 2H )* it is probable that D-1

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IEED0-24782 ,

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               -    the existing governor system, with its present calibration, would not be capable of controlling the acceleration transient of the BPC1 turbine. The turbine I

would probably trip due to mechanical overspeed. Bowever, it should be noted

                                                                            ~

that the mechanical overspeed trip device ta capable'of automatic rerat. , Therefore, although not. satisfying the system startup ' time criterion of 25 ses - 1 onds, the turbine (and system) is capable of restarting and ultimately deliver-ing design flow rate, even with high concentrations of hydrogen gas. l Bydrogen and steam will vent from the turbine during operation because the gland seal system will not innetion as designed,due to high concent. ration of non-condensible hydrogen gas. This would not preclude turbine operation, but should be considered if the system were ever operated in this mode, due to the obvious hazards of fire and radiation in the reactor building. L I l'

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